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It is well documented that the central electron temperature in the national spherical torus experiment (NSTX) remains largely unchanged as the external heating power, and hence the normalized volume averaged plasma pressure β increases [Stutman, Phys. Rev. Lett. 102, 115002 (2009)PRLTAO0031-900710.1103/PhysRevLett.102.115002]. Here we present a hypothesis that low n, pressure driven ideal magnetohydrodynamic (MHD) instabilities that are nondisruptive, can break magnetic surfaces in the central region and thereby flatten the electron temperature profiles. We demonstrate this mechanism in a 3D resistive MHD simulation of a NSTX discharge. By varying the toroidal magnetic field strength, and/or the heating power, we show that there is a critical value of β, above which the central temperature profile no longer peaks on axis.

Beam modulation effects on Alfvén eigenmode stability have been investigated in a recent DIII-D experiment and show that variations in neutral beam modulation period can have an impact on the beam driven Alfvén eigenmode spectrum and resultant fast ion transport despite similar time-averaged input power. The experiment was carried out during the current ramp phase of L-mode discharges heated with sub-Alfvénic 50–80 kV deuterium neutral beams that drive a variety of Alfvén eigenmodes unstable. The modulation period of two interleaved beams with different tangency radii was varied from shot to shot in order to modify the relative time dependent mix of the beam pitch angle distribution as well as the persistence of a bump-on-tail feature near the injection energy (a feature confirmed by imaging neutral particle analyzer measurements). As the beam modulation period is varied from 7 ms to 30 ms on/off (typical full energy slowing down time of τ slow ≈ 50 ms at mid-radius), toroidicity-induced Alfvén eigenmodes (TAEs) located in the outer periphery of the plasma become intermittent and coincident with the more tangential beam. Core mode activity changes from reversed shear Alfvén eigenmodes (RSAEs) to a mix of RSAE and beta-induced Alfvén eigenmodes. Discharges with 30 ms on/off period do not have a persistent bump-on-tail feature, have the lowest average mode amplitude and least fast ion transport. Detailed analysis of an individual TAE using TRANSP kick modeling (Monte Carlo evolution of the distribution function with probabilistic ‘kicks’ by the AEs) and the resistive MHD code with kinetic fast ions, MEGA, find no strong role of energy gradient drive due to bump-on-tail features. Instead, the observed TAE modulation with interleaved beams is likely a pitch angle dependent result combined with slowing down of the tangential beam between pulses. For the conditions investigated, bump-on-tail contributions to TAE drive were found to be 5% or less of the total drive at any given time.

The mission of the low aspect ratio spherical tokamak NSTX-U is to advance the physics basis and technical solutions required for optimizing the configuration of next-step steady-state tokamak fusion devices. NSTX-U will ultimately operate at up to 2 MA of plasma current and 1 T toroidal field on axis for 5 seconds, and has available up to 15 MW of Neutral Beam Injection (NBI) power at different tangency radii and 6 MW of High Harmonic Fast Wave (HHFW) heating. With these capabilities NSTX-U will develop the physics understanding and control tools to ramp-up and sustain high performance fully non-inductive plasmas with large bootstrap fraction and enhanced confinement enabled via the low aspect ratio, high beta configuration. With its unique capabilities, NSTX-U research also supports ITER and other critical fusion development needs. Super-Alfvénic ions in beam-heated NSTX-U plasmas access energetic particle parameter space that is relevant for both -heated conventional and low aspect ratio burning plasmas. NSTX-U can also generate very large target heat fluxes to test conventional and innovative plasma exhaust and plasma facing component (PFC) solutions. This paper summarizes recent analysis, theory and modelling progress to advance the tokamak physics basis in the areas of macrostability and 3D fields, energetic particle stability and fast ion transport, thermal transport and pedestal structure, boundary and plasma material interaction, RF heating, scenario optimization and real-time control.

The structure of the non-axisymmetric heat load distribution at the divertor plates is determined not only by the toroidal but also from the poloidal spectrum of non-axisymmetric field perturbations. Whether they are intrinsic, like error fields, or they are applied through 3D coils, the non-axisymmetric fields produce complex 3D edge magnetic topologies (footprints) that alter the properties of the heat and particle flux distributions on the divertor target plates. In this manuscript, a study of the impact of applied 3D field poloidal spectrum on the footprint size and structure is done for the DIII-D tokamak using the resistive MHD code M3D-C1 coupled with the field line tracing code TRIP3D. To resolve the impact of the poloidal spectrum of the magnetic perturbation, the relative phase of the two rows of in-vessel 3D coils used to produce both a n = 2 and a n = 3 perturbation is varied, where n is the toroidal harmonic of the magnetic perturbation. This shows that the largest footprint is predicted when the relative phase of the two rows is close to zero, which is also where the resonant coupling with the plasma is maximized. These results suggest that it will be challenging to decouple the footprint size from the requisite resonant coupling for RMP–ELM control. The correlation between the measured heat load and particle flux distributions at the outer divertor plates in DIII-D and the magnetic measurements is in good agreement with the predicted dependence of the magnetic footprint size on the amplitude of the resonant component of the plasma response.

This paper presents a multi-machine, multi-parameter scaling law for the $n=2$ core resonant error field threshold that leads to field penetration, locked modes, and disruptions. Here, $n$ is the toroidal harmonic of the non-axisymmetric error field (EF). While density scalings have been reported by individual tokamaks in the past, this work performs a regression across a comprehensive range of densities, toroidal fields, and pressures accessible across three devices using a common metric to quantify the EF in each device. The metric used is the amount of overlap between an EF and the spectrum that drives the largest linear ideal MHD resonance, known as the "dominant mode overlap". This metric, which takes into account both the external field and plasma response, is scaled against experimental parameters known to be important for the inner layer physics. These scalings validate nonlinear MHD simulation scalings, which are used to elucidate the dominant inner layer physics. Both experiments and simulations show that core penetration thresholds for EFs with toroidal mode number $n=2$ are of the same order as the $n=1$ thresholds that are considered most dangerous on current devices. Both $n=1$ and $n=2$ thresholds scale to values within the ITER design tolerances, but data from additional devices with a range of sizes are needed in order to increase confidence in quantitative extrapolations of $n=2$ thresholds to ITER.

Optimizing the differential pair connection scheme (i.e., the set of pairs) of a toroidal array of magnetic sensors dedicated to measuring slowly rotating asymmetric fields can enhance the mode number detection capability and failure-resilience. In this work, the condition number obtained from singular value decomposition of the design matrix is used as a metric to evaluate the quality of a connection scheme. A large number of possible pair connections are usually available, so evaluating all of them may require extensive use of computational resources and can be very time-consuming. Alternative methods to reduce the number of pairs evaluated without losing the capabilities of toroidal mode detection are presented in this paper. Three examples of the applications of such analysis for the 3D magnetic diagnostic system of DIII-D are also presented: the addition of two new toroidal arrays with n > 3 detection capabilities, the modification of an existing toroidal array in the low field side of the machine to accommodate the addition of a helicon antenna, and the design of changes in several toroidal arrays in the high field side to accommodate the addition of a lower hybrid current drive antenna on the center post.

The successful application of three-dimensional (3D) magnetohydrodynamic (MHD) spectroscopy enables us to identify the multi-mode eigenvalues in DIII-D and KSTAR tokamak experiments with stable plasmas. The temporal evolution of the multi-modes’ stabilities have been detected. The new method is numerically efficient allowing the real time detection of MHD modes’ stabilities during the discharge. The method performs active detection of the plasma stability by utilizing the upper and lower rows of internal non-axisymmetric coils to apply a wide variety of 3D fields. Multi-mode eigenvalues are extracted using subspace system identification of the plasma response measured by 3D-field magnetic sensors distributed at different poloidal locations. The equivalence of this new method with the one introduced by Wang (2019 Nucl. Fusion 59 024001) has been numerically corroborated. The more robust and efficient calculation developed here will enable real time monitoring of the plasma stability based on the extracted eigenvalues of stable modes.

Given spatially resolved measurements of normal and tangential components of the magnetic field just outside the surface of a magnetically confined plasma, the field at the measurement location can be uniquely decomposed into contributions from the plasma and from external sources. This principle allows direct measurement of the electromagnetic torque on the plasma without knowledge of the distribution of the internal and external currents, similar to the more well-known formalism using the Maxwell stress tensor. The internal/external field decomposition also enables a mixed approach that incorporates any explicitly known current distributions (e.g., from non-axisymmetric coils). We discuss the requirements and limitations of such an approach to torque measurements. Experimental measurements of the torque evolution as a rotating tearing mode locks to the wall in the DIII-D tokamak are consistent with a simple model.

Alfvén Eigenmodes (AE) and magneto-hydrodynamic (MHD) modes are destabilized in DIII-D reverse magnetic shear configurations and may limit the performance of the device. We use the reduced MHD equations in a full 3D system, coupled with equations of density and parallel velocity moments for the energetic particles (with gyro-fluid closures) as well as the geodesic acoustic wave dynamics, to study the properties of instabilities observed in DIII-D reverse magnetic shear discharges. The aim of the study consists in finding ways to avoid or minimize MHD and AE activity for different magnetic field configurations and neutral beam injection (NBI) operational regimes. The simulations show at the beginning of the discharge, before the reverse shear region is formed, a plasma that is AE unstable and marginally MHD stable. As soon as the reverse shear region appears, ideal MHD modes are destabilized with a larger growth rate than the AEs. Both MHD modes and AEs coexist during the discharge, although the MHD modes are more unstable as the reverse shear region deepens. The simulations indicate the destabilization of Beta induced AE (BAE), Toroidal AE (TAE), elliptical AE (EAE) and reverse shear AE (RSAE) at different phases of the discharges, showing a reasonable agreement between the frequency range of the dominant modes in the simulations and the diagnostic measurements (...) This material based on work is supported both by the U.S. Department of Energy, Office of Science, under Contract DE-AC05-00OR22725 with UT-Battelle, LLC and U.S. Department of Energy, Oce of Science, Oce of Fusion Energy Sciences, using the DIII-D National Fusion Facility, a DOE Oce of Science user facility, under Award No. DE-FC02-04ER54698. This research was sponsored in part by the Ministerio of Economía y Competitividad of Spain under project no.ENE2015-68265-P. DIII-D data shown in this paper can be obtained in digital format by following the links at https://fusion.gat.com/global/D3D DMP. Publicado

The presence of error fields in a tokamak device is inevitable, so it is of fundamental importance to understand their impact on the operation of the device and up to what level they can be tolerated. In this paper a prediction of the impact on the magnetic footprints on the divertor plates due to the misalignment of NSTX-U equilibrium coils is presented. Resistive MHD simulations are used to predict the magnetic field perturbations experienced by the field lines due to the presence of error fields. A linear relation between the magnitude of the misalignment and the area of the magnetic footprints is found, as well as an inverse proportionality between the size of a magnetic footprint and the plasma penetration of the field lines composing the footprint. This study is intended as a starting point in the process of investigating divertor footprints associated with the plasma response to intrinsic error fields with resistive MHD codes. A basic interpretation of the results suggests that large footprints resulting from equilibrium coils misalignment may be desirable, to some extent, to decrease the peak heat load on the divertor components. A more complete interpretation of the results presented here will involve a number of complex physics issues, for example pedestal transport and stability, along with the role of various scrape-off layer and divertor processes, as well as experimental data to confirm the predictions.

In the RFX-mod Reversed Field Pinch experiment the wall is entirely covered by graphite tiles. The disadvantage of a graphite wall is that, except after wall conditioning procedures, such as boronization, lithization or glow discharge cleaning, the graphite becomes saturated with trapped Hydrogen, causing that during the discharge the particle recycling factor is approximately one and plasma density is entirely sustained by particle fluxes coming from the wall. In the paper the results of the analyses of graphite samples after their exposure to plasmas or wall conditioning treatments are presented. The topics addressed are: the deposition of Boron during different boronization treatments; the erosion and re-deposition of the Boron layer by Hydrogen plasmas; some remarks about the presence of Lithium in the wall since the beginning of lithization treatments.

Adding toroidal arrays of magnetic probes at the top and bottom of NSTX-U would improve both the detection of the multimodal plasma response to applied magnetic perturbations and the identification of the poloidal structure of unstable plasma modes, as well as contribute to the validation of MHD models, improve the understanding of the plasma response to external fields, and improve the error field correction. In this paper, the linear MHD code MARS-F/K has been used to identify poloidal locations that would improve the capability to measure stationary or near-stationary 3D fields that may result from the plasma response to external sources of non-axisymmetric fields. The study highlighted 6 poloidal positions where new arrays of both poloidal and radial magnetic field sensors would improve the poloidal resolution. The proposed set of new arrays combined with the present ones is shown to be capable of measuring the poloidal structure of perturbations with n ≤ 6 and of detecting the multimodal plasma response. Assessment of the trade-off in the poloidal length of the probes leads to an ideal length between 10 cm and 30 cm. A method to configure the probes of a toroidal array based on the singular value decomposition condition number is proposed, and an ideal solution and a low-cost one are presented.

Previous DIII-D research has identified the presence of a multimodal plasma response to perturbations with toroidal periodicity equal to 2, with kink-like phenomenology of the plasma response in the low field side mid-plane and a resonant-like behavior in the high field side midplane. In this paper, we extend that initial midplane observation with the first detailed analysis of the poloidal structure of the plasma response as a function of the poloidal spectrum of the applied n = 2 perturbation and compare it to predictions made with the linear magnetohydrodynamics code MARS-F, confirming an overall good agreement. Singular value decomposition analysis of both the experimental data and the modeling identifies two distinct poloidal mode structures, one peaking at the low field side midplane and one peaking at θ ± 50°, each with a well-defined dependence on the poloidal spectrum of the applied perturbation. A study of the effect that βN has on the two modes confirms the kink-like phenomenology of the plasma response that dominates the low field side mid-plane, shows that the behavior previously associated with the high field side is observable at all poloidal angles, and highlights that the mode dominant at low βN becomes the secondary one at high βN. Detailed analysis of the high field side response is presented as well, suggesting the presence of a third mode linked to the presence of non-linear effects corresponding to particular poloidal spectra of the applied perturbation.Previous DIII-D research has identified the presence of a multimodal plasma response to perturbations with toroidal periodicity equal to 2, with kink-like phenomenology of the plasma response in the low field side mid-plane and a resonant-like behavior in the high field side midplane. In this paper, we extend that initial midplane observation with the first detailed analysis of the poloidal structure of the plasma response as a function of the poloidal spectrum of the applied n = 2 perturbation and compare it to predictions made with the linear magnetohydrodynamics code MARS-F, confirming an overall good agreement. Singular value decomposition analysis of both the experimental data and the modeling identifies two distinct poloidal mode structures, one peaking at the low field side midplane and one peaking at θ ± 50°, each with a well-defined dependence on the poloidal spectrum of the applied perturbation. A study of the effect that βN has on the two modes confirms the kink-like phenomenology of the plasma...

In the reversed field pinch RFX-mod at the highest plasma current of 2 MA, when error fields are not effectively feedback controlled, localized thermal loads up to tens of MW m-2 can be produced. The graphite tiles withstand such high power loads, but the high hydrogen retention makes density control extremely difficult. Several wall conditioning techniques have been optimized in the last campaigns, including helium glow discharge cleaning and wall boronization by diborane glow discharges. More recently, lithium conditioning has been applied for the first time in a reversed field pinch by the evaporation technique. The main results are discussed in this paper. Lithization leads to important operational advantages: a significant improvement of the density control is obtained. Densities up to n/nG ≈ 0.5 can be produced in a controlled way. At the same value of input power, plasmas at higher densities can be sustained. However, due to the short particle confinement time, such densities are reached with high rates of gas puffing and the resulting profiles at high density are edge peaked. A lithium multipellet injector, to be applied in order to obtain a more uniform deposition, has been tested. © 2013 IAEA, Vienna.

The orientation of 3D equilibria in the Madison Symmetric Torus (MST) [R. N. Dexter et al., Fusion Technol. 19, 131 (1991)] reversed-field pinch can now be controlled with a resonant magnetic perturbation (RMP). Absent the RMP, the orientation of the stationary 3D equilibrium varies from shot to shot in a semi-random manner, making its diagnosis difficult. Produced with a poloidal array of saddle coils at the vertical insulated cut in MST's thick conducting shell, an m = 1 RMP with an amplitude br/B ∼ 10% forces the 3D structure into any desired orientation relative to MST's diagnostics. This control has led to improved diagnosis, revealing enhancements in both the central electron temperature and density. With sufficient amplitude, the RMP also inhibits the generation of high-energy (>20 keV) electrons, which otherwise emerge due to a reduction in magnetic stochasticity in the core. Field line tracing reveals that the RMP reintroduces stochasticity to the core. A m = 3 RMP of similar amplitude has little...

This paper summarizes the main achievements of the RFX fusion science program in the period between the 2008 and 2010 IAEA Fusion Energy Conferences. RFX-mod is the largest reversed field pinch in the world, equipped with a system of 192 coils for active control of MHD stability. The discovery and understanding of helical states with electron internal transport barriers and core electron temperature >1.5 keV significantly advances the perspectives of the configuration. Optimized experiments with plasma current up to 1.8 MA have been realized, confirming positive scaling. The first evidence of edge transport barriers is presented. Progress has been made also in the control of first-wall properties and of density profiles, with initial first-wall lithization experiments. Micro-turbulence mechanisms such as ion temperature gradient and micro-tearing are discussed in the framework of understanding gradient-driven transport in low magnetic chaos helical regimes. Both tearing mode and resistive wall mode active control have been optimized and experimental data have been used to benchmark numerical codes. The RFX programme also provides important results for the fusion community and in particular for tokamaks and stellarators on feedback control of MHD stability and on three-dimensional physics. On the latter topic, the result of the application of stellarator codes to describe three-dimensional reversed field pinch physics will be presented. © 2011 IAEA, Vienna.

The full three-dimensional (3D) approach is now becoming an important issue for all magnetic confinement configurations. It is a necessary condition for the stellarator but also the tokamak and the reversed field pinch (RFP) now cannot be completely described in an axisymmetric framework. For the RFP the observation of self-sustained helical configurations with improved plasma performances require a better description in order to assess a new view on this configuration. In this new framework plasma configuration studies for RFX-mod have been considered both with tools developed for the RFP as well as considering codes originally developed for the stellarator and adapted to the RFP. These helical states are reached through a transition to a very low/reversed shear configuration leading to internal electron transport barriers. These states are interrupted by MHD reconnection events and the large Te gradients at the barriers indicate that both current and pressure driven modes are to be considered. Furthermore the typically flat Te profiles in the helical core have raised the issue of the role of electrostatic and electromagnetic turbulence in these reduced chaos regions, so that a stability analysis in the correct 3D geometry is required to address an optimization of the plasma setup. In this viewtheVMECcode proved to be an effectiveway to obtain helical equilibria to be studied in terms of stability and transport with a suite of well tested codes. In this work, the equilibrium reconstruction technique as well as the experimental evidence of 3D effects and their first interpretation in terms of stability and transport are presented using both RFP and stellarator tools. © 2010 IOP Publishing Ltd.

Single Helical Axis (SHAx) states obtained in high current reversed field pinch (RFP) plasmas display, aside from a dominant mode in the m=1 spectrum, also a dominant m=0 mode, with the same toroidal mode number as the m=1 one. The two modes have a fixed phase relationship. The island chain created by the m=0 mode across the reversal surface gives rise, at shallow reversal of the toroidal field, to an X-point structure which separates the last closed flux surface from the first wall, creating a divertor-like configuration. The plasma-wall interaction is found to be related to the connection length of the field lines intercepting the wall, which displays a pattern modulated by the dominant mode toroidal periodicity. This configuration, which occurs only for shallow toroidal field reversal, could be exploited to realize an island divertor in analogy to stellarators., 12 pages, 9 figures Submitted to Nuclear Fusion

The reversed field pinch configuration is characterized by the presence of magnetic structures both in the core and at the edge: in the core, at high plasma current the spontaneous development of a helical structure is accompanied by the appearance of internal electron transport barriers; at the edge strong pressure gradients, identifying an edge transport barrier, are observed too, related to the position of the field reversal surface. The aim of this paper is the experimental characterization of both the internal and edge transport barriers in relation to the magnetic topology, discussing possible analogies and differences with other confinement schemes. © 2009 IOP Publishing Ltd.

In this contribution a detailed study of the edge flow properties performed on the outer region of the RFX-mod experiment using a Gundestrup probe is presented. In particular, both parallel and perpendicular components (with respect to the magnetic field) have been evaluated in several low current discharges providing a complete reconstruction of the edge flow profiles. Different methods based on both magnetized and unmagnetized models have been used to evaluate the averaged plasma flow components and their time variations. Furthermore, configurations involving either ion saturation current signals or floating potential ones have been used to evaluate both parallel and perpendicular flow components and a good agreement has been found between different methods. Finally, the dynamic behaviour of edge flow fluctuations induced by gradient modifications following pellet injection and by varying the edge magnetic topology (the m = 0 islands radial position) have been studied (© 2010 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim)

To aid in diagnosis of 3D equilibria in the Madison Symmetric Torus, it has become necessary to control the orientation of the equilibria. In reversed field pinch experiments a transition to a 3D equilibrium is common with sufficiently large plasma current (and Lundquist number). Diagnosis of this state is hampered by the fact that the helical structure is stationary but with an orientation that varies shot-to-shot. A resonant magnetic perturbation (RMP) technique has been developed to vary controllably the orientation of the 3D equilibria and optimized to minimize the plasma wall interaction due to its use. Application of an RMP now allows alignment of the structure with key diagnostics, including Thomson scattering and an interferometer-polarimeter.

This paper reports the highlights of the RFX-mod fusion science programme since the last 2010 IAEA Fusion Energy Conference. The RFX-mod fusion science programme focused on two main goals: exploring the fusion potential of the reversed field pinch (RFP) magnetic configuration and contributing to the solution of key science and technology problems in the roadmap to ITER. Active control of several plasma parameters has been a key tool in this endeavour. New upgrades on the system for active control of magnetohydrodynamic (MHD) stability are underway and will be presented in this paper. Unique among the existing fusion devices, RFX-mod has been operated both as an RFP and as a tokamak. The latter operation has allowed the exploration of edge safety factor q edge < 2 with active control of MHD stability and studies concerning basic energy and flow transport mechanisms. Strong interaction has continued with the stellarator community in particular on the physics of helical states and on three-dimensional codes. © 2013 IAEA, Vienna.

Plasma–wall interaction is one of the most important issues that present magnetic confinement devices have to face. In the RFX-mod reversed field pinch experiment plasma–wall interaction has become a hard point increasing plasma current up to the RFX-mod maximum design value of 2 MA, since in this case local power deposition can be as high as 10 MW m−2. Since the first wall of RFX-mod is entirely covered by graphite tiles different techniques have been tested to control hydrogen wall influx: He glow discharges cleaning, He discharges at high plasma currents, wall boronization and baking. With the best results obtained by boronization, at high plasma currents all such techniques improve the situation but do not allow a complete and stationary hydrogen influx reduction. Furthermore, in the presence of localized high power load the wall still responds providing very high influxes. In order to improve this situation wall conditioning by lithium has been tested. As a first lithization method to deposit a controllable amount of lithium on the wall, a room temperature pellet injector has been used (maximum pellet diameter of 1.8 mm and maximum length of 5 mm). Lithium coatings with a theoretical thickness of about 10 nm have been applied both to clean graphite tiles and over boronized ones. Lithization demonstrated to be effective in lowering hydrogen wall recycling to a value smaller than that of boronized graphite, with the effect lasting 20–30% more than in the boronized case. Compared with boronization, lithization slightly improves (by about 30%) particle confinement time and also clearly affects edge particle transport providing a lower edge density and more peaked density profiles. Lithization also reduces carbon content by about 10% over boronization but still no clear improvement has been observed in terms of energy confinement. Similar results have been obtained performing lithization over boronized graphite.

To evaluate the effect of the prostaglandin inhibitor acetylsalicylic acid (ASA) on rat exocrine pancreas secretion, three groups of rats were administered ASA by infusion: Groups 1-3, 50, 100, and 200 mg/kg body wt, respectively; Group 4 received saline. Twenty minutes later these ASA-pretreated groups were given intraarterial secretin (18 CU/kg) and cholecystokinin (CCK) (18 micrograms/kg). In an additional three groups of seven rats each, saline solution rather than secretin-CCK was given after ASA pretreatment. Pancreatic juice was collected every 10 min by means of a chronic pancreatic fistula. Bicarbonate and protein concentrations were measured and variations in outputs observed. No significant variations were found in the bicarbonate concentrations and outputs of rats with different types of pharmacological treatment, while protein concentrations and outputs were found to vary with time and type of experiment. There was, however, no interaction between these two variables. At lower ASA dosages, the bicarbonate and protein concentrations and outputs of secretin-CCK-stimulated rats were higher than the basal values and the levels of rats without hormonal stimulation. At higher dosages, no difference was found between the two groups. In conclusion, ASA seems to interfere with stimulated pancreatic exocrine secretion of proteins, even when its effect on bicarbonate concentration is factored in, and its effect seems to be present at the highest dosages considered in the study. Among the various hypotheses that may explain this phenomenon, an antagonizing effect of ASA on secretin-CCK action should be the first to be considered.

The transport of injected impurities is investigated in the multiple helicity (MH) and improved confinement quasi single helicity (QSH) magnetic regimes of RFX-mod reversed-field pinch. Impurities are introduced through injections of room temperature pellets of carbon (in MH and QSH) and lithium (in MH). Simulation of experimental spectroscopic measurements, including line and soft x-rays emission, electron density and radiated power is carried out with a 1D collisional-radiative impurity transport code, allowing to determine the transport parameters: diffusion coefficient and convective velocity. The transport coefficients were found close to those deduced in previous experiments through the introduction of Ni and Ne, confirming the existence of an external velocity barrier also for light impurities.

This paper presents the subtleties of obtaining robust experimental scaling laws for the core resonant error field threshold that leads to field penetration, locked modes, and disruptions. Recent progress in attempts to project this threshold to new machines has focused on advances in the metric used to quantify the dangerous error fields, incorporating the ideal MHD plasma response in a metric referred to as the 'dominant mode overlap'. However, the scaling of this or any quantity with experimental parameters known to be important for the complicated tearing layer physics requires regressions performed for databases that, for historical reasons, unevenly sample the available parametric space. This paper presents the distribution of the existing international n = 1 database and details biases in the available sampling and details the sensitivity of ITER projections to simple least-squares regressions. Downsampling and a simple kernel density estimation weighted regression are used here to demonstrate the difference in projections that acknowledging the machine sampling bias can make. This results in more robust projection to parameters far from the 'usual' devices built thus far. Two multi-device and multi-parameter scalings of the EF threshold in Ohmic and powered plasmas are presented, projecting the threshold to ITER and investigating the impact of sampling biases.

In normal operation, Madison Symmetric Torus (MST) [R. N. Dexter et al., Fusion Technol. 19, 131 (1991)] reversed-field pinch plasmas exhibit several rotating tearing modes (TMs). Application of a resonant magnetic perturbation (RMP) results in braking of mode rotation and, if the perturbation amplitude is sufficiently high, in a wall-locked state. The coils that produce the magnetic perturbation in MST give rise to RMPs with several toroidal harmonics. As a result, simultaneous deceleration of all modes is observed. The measured TM dynamics is shown to be in qualitative agreement with a magnetohydrodynamical model of the RMP interaction with the TM [R. Fitzpatrick, Nucl. Fusion 33, 1049 (1993)] adapted to MST. To correctly model the TM dynamics, the electromagnetic torque acting on several TMs is included. Quantitative agreement of the TM slowing-down time was obtained for a kinematic viscosity in the order of νkin≈10–20 m2/s. Analysis of discharges with different plasma densities shows an increase of th...