ITER Scenarios Research Articles

Detecting neoclassical tearing modes in high-temperature ITER plasma scenarios with the ITER prototype electron cyclotron emission diagnostic

Abstract Electron cyclotron emission (ECE) diagnostics for ITER serve two key purposes. The diagnostics will measure plasma electron temperature with high spatial and temporal resolution. Additionally, they will be used to detect neoclassical tearing modes (NTMs), a deleterious and nonlinearly unstable mode causing the growth of magnetic ‘seed’ islands. Interpreting ECE requires anticipation of physical limits including frequency cut-offs and harmonic overlap. In high temperature plasmas, the relativistic shift and broadening of the emission must also be considered to accurately reconstruct the electron temperature spatial profile. Accounting for these effects allows ECE diagnostics to be used for accurate measurement of the equilibrium electron temperature profile, as well as fluctuations about this equilibrium. One such fluctuation is caused by the fast radial transport of heat across rotating magnetic islands. ECE diagnostics can detect this change as an oscillation at the plasma rotation frequency to determine the existence and location of NTMs. This paper presents work on a synthetic diagnostic for ECE. The synthetic diagnostic tests simulated ECE signals, which are inferred from ITER scenarios perturbed by magnetic islands after accounting for all ECE physics. The synthetic diagnostic tests conventional ECE detection algorithms for NTMs in real-time on ITER-recommended hardware. Combined, these two areas of focus help determine design of the ECE system.

Quantitative phase imaging verification in large field-of-view lensless holographic microscopy via two-photon 3D printing

Large field-of-view (FOV) microscopic imaging (over 100 mm2) with high lateral resolution (1–2 μm) plays a pivotal role in biomedicine and biophotonics, especially within the label-free regime. Lensless digital holographic microscopy (LDHM) is promising in this context but ensuring accurate quantitative phase imaging (QPI) in large FOV LDHM is challenging. While phantoms, 3D printed by two-photon polymerization (TPP), have facilitated testing small FOV lens-based QPI systems, an equivalent evaluation for lensless techniques remains elusive, compounded by issues such as twin-image and beam distortions, particularly towards the detector’s edges. Here, we propose an application of TPP over large area to examine phase consistency in LDHM. Our research involves fabricating widefield phase test targets with galvo and piezo scanning, scrutinizing them under single-shot twin-image corrupted conditions and multi-frame iterative twin-image minimization scenarios. By measuring the structures near the detector’s edges, we verified LDHM phase imaging errors across the entire FOV, with less than 12% phase value difference between areas. Our findings indicate that TPP, followed by LDHM and Linnik interferometry cross-verification, requires new design considerations for precise large-area photonic manufacturing. This research paves the way for quantitative benchmarking of large FOV lensless phase imaging, enhancing understanding and further development of LDHM technique.

Ripple-induced neoclassical toroidal viscous torque in Augmented-First Plasma operation phase in ITER

Abstract A systematic calculation is performed on the ripple-induced neoclassical toroidal viscous (NTV) torque for new ITER scenarios designed for the Augmented-First Plasma (A-FP) operation phase with the full tungsten wall, where the plasma-wall gap is varied in view of mitigating the impact of tungsten wall-plasma interactions. The torque calculation includes drift kinetic response of the plasma thermal and energetic particles to the n = 18 (n is the toroidal harmonic number) ripple field. For the plasma scenario with ~45 cm plasma-wall gap at the outboard mid-plane and considering the corrected ripple level of 0.17% by the ferritic steel inserts, the computed net NTV torque acting on the plasma column is in the sub-Nm level. However, with decreasing the plasma-wall gap, the computed net NTV torque can reach a level comparable to that produced by the neutral-beam momentum injection in ITER. Ripple correction by ferritic inserts reduces the net torque by a factor of 3.3 for all the three A-FP scenarios considered. The n ω d = l ω b (with ω d and ω b being the toroidal precession and bounce frequencies of trapped particles, respectively, and l an integer number) type of resonance-enhancement of the NTV torque, due to thermal particles, is found to be weak in ITER despite high-n of 18. The same also holds for the ITER 10 MA steady state scenario from the D-T operation phase, where the aforementioned resonance associated with fusion-born alphas is also included. The ripple-induced NTV torque is well below that produced by the resonant magnetic perturbation applied for controlling the type-I edge-localized mode in ITER.

DIII-D research to provide solutions for ITER and fusion energy

The DIII-D tokamak has elucidated crucial physics and developed projectable solutions for ITER and fusion power plants in the key areas of core performance, boundary heat and particle transport, and integrated scenario operation, with closing the core-edge integration knowledge gap being the overarching mission. New experimental validation of high-fidelity, multi-channel, non-linear gyrokinetic turbulent transport models for ITER provides strong confidence it will achieve Q ⩾ 10 operation. Experiments identify options for easing H-mode access in hydrogen, and give new insight into the isotopic dependence of transport and confinement. Analysis of 2,1 islands in unoptimized low-torque IBS demonstration discharges suggests their onset time occurs randomly in the constant β phase, most often triggered by non-linear 3-wave coupling, thus identifying an NTM seeding mechanism to avoid. Pure deuterium SPI for disruption mitigation is shown to provide favorable slow cooling, but poor core assimilation, suggesting paths for improved SPI on ITER. At the boundary, measured neutral density and ionization source fluxes are strongly poloidally asymmetric, implying a 2D treatment is needed to model pedestal fuelling. Detailed measurements of pedestal and SOL quantities and impurity charge state radiation in detached divertors has validated edge fluid modelling and new self-consistent ‘pedestal-to-divertor’ integrated modeling that can be used to optimize reactors. New feedback adaptive ELM control minimizes confinement reduction, and RMP ELM suppression with sustained high core performance was obtained for the first time with the outer strike point in a W-coated, compact and unpumped small-angle slot divertor. Advances have been made in integrated operational scenarios for ITER and power plants. Wide pedestal intrinsically ELM-free QH-modes are produced with more reactor-relevant conditions, Low torque IBS with W-equivalent radiators can exhibit predator-prey oscillations in T e and radiation which need control. High-β P scenarios with q min > 2, q 95–7.9, β N > 4, β T–3.3% and H 98y2 > 1.5 are sustained with high density ( n¯ = 7E19 m−3, f G–1) for 6 τ E, improving confidence in steady-state tokamak reactors. Diverted NT plasmas achieve high core performance with a non-ELMing edge, offering a possible highly attractive core-edge integration solution for reactors.

Resistive wall mode and fishbone mode in ITER steady state scenario: roles of fusion-born alphas and plasma flow

Drift-kinetic effects of fusion-born alpha particles on the n= 1 (n is the toroidal mode number) resistive wall mode (RWM) is numerically investigated for a recent design of the ITER 10 MA steady state plasma scenario, utilizing a magneto-hydrodynamic (MHD)-kinetic hybrid toroidal model. While the fluid theory predicts unstable RWM as the normalized plasma pressure β N exceeds the no-wall Troyon limit and with the mode growth rate monotonically increasing with β N, inclusion of the drift-kinetic contribution of trapped alphas qualitatively modifies the behavior by stabilizing the mode at high β N. In fact, a complete stabilization of the n= 1 RWM up to the ideal-wall Troyon limit is found. On the other hand, another unstable branch—the alpha-driven n = 1 fishbone mode (FB)—is identified in the high-β N regime, with the mode frequency matching that of the toroidal precession frequency of trapped alphas. Fast plasma toroidal flow however helps mitigate the FB instability. Kinetic stabilization of the RWM and flow stabilization of the (alpha-triggered) FB result in an enhancement of β N from the design value of 3.22–3.52 for the ITER scenario considered, while still maintaining stable plasma operation against the aforementioned MHD instabilities.

Fluid and kinetic studies of tokamak disruptions using Bayesian optimization

When simulating runaway electron dynamics in tokamak disruptions, fluid models with lower numerical cost are often preferred to more accurate kinetic models. The aim of this work is to compare fluid and kinetic simulations of a large variety of different disruption scenarios in ITER. We consider both non-activated and activated scenarios; for the latter, we derive and implement kinetic sources for the Compton scattering and tritium beta decay runaway electron generation mechanisms in our simulation tool Dream (Hoppe et al., Comput. Phys. Commun., vol. 268, 2021, 108098). To achieve a diverse set of disruption scenarios, Bayesian optimization is used to explore a range of massive material injection densities for deuterium and neon. The cost function is designed to distinguish between successful and unsuccessful disruption mitigation based on the runaway current, current quench time and transported fraction of the heat loss. In the non-activated scenarios, we find that fluid and kinetic disruption simulations can have significantly different runaway electron dynamics, due to an overestimation of the runaway seed by the fluid model. The primary cause of this is that the fluid hot-tail generation model neglects superthermal electron transport losses during the thermal quench. In the activated scenarios, the fluid and kinetic models give similar predictions, which can be explained by the significant influence of the activated sources on the runaway dynamics and the seed.

Prediction of pellet mass thresholds for ELM triggering in low-collisionality, ITER-like discharges

In ITER, pellets are calculated to require more than 8 times the mass than currently planned to reliably trigger edge-localized modes (ELMs). Unmitigated heat flux impulses from ELMs are intolerable in ITER at full power and current. Therefore, ITER operation relies on multiple approaches to control ELM heat fluxes. One method is pellet ELM pacing to instigate small rapid ELMs with low heat flux. Predicting the performance of pellet pacing is critical for ITER, which is expected to operate in a regime with a low-collisionality, peeling-limited pedestal. However, to trigger ELMs the local pressure increase in the expanding pellet cloud pushes the equilibrium over the ballooning stability limit. In this work, linear and nonlinear M3D-C1 simulations are used to predict pellet mass thresholds in DIII-D discharges and ITER scenarios with peeling-limited pedestals. It is found that the distance of the equilibrium’s operational point from the ballooning branch of the pedestal stability boundary strongly changes thresholds. Linear M3D-C1 simulations find a strong dependence of the pellet mass threshold on the poloidal injection location for ITER’s 15 MA, Q = 10 scenario. The required pellet mass at the planned injection locations is 8 to 17 times larger than currently considered. However, such linear simulations do not include pellet ablation physics or time evolution of density and temperature. A new scheme of 2D nonlinear simulations, coupled with linear stability analysis at various steps throughout the nonlinear time evolution, was developed to include such physics and improve on the linear results. These new nonlinear-to-linear simulations confirm previous findings. This result suggests that pellet ELM triggering in ITER could require pellets much larger than those currently planned, which makes ELM-pacing operationally challenging. On the other hand, fueling pellets injected from the high-field side will likely not unintentionally trigger ELMs in an otherwise ELM-stable plasma.

Integrated analysis of plasma rotation effect on HL-3 hybrid scenario

Abstract The hybrid scenario, which has good confinement and moderate MHD instabilities, is a proposed operation scenario for international thermonuclear experimental reactor (ITER). In this work, the effect of plasma rotation on the HL-3 hybrid scenario is analyzed with the integrated modeling framework OMFIT. The results show that toroidal rotation has no obvious effect on confinement with a high line averaged density of n bar ∼ 7 × 1019 m−3. In this case, the ion temperature only changes from 4.7 keV to 4.4 keV with the rotation decreasing from 105 rad/s to 103 rad/s, which means that the turbulent heat transport is not dominant. While in the scenarios characterized by lower densities, such as n bar ∼ 4 × 1019 m−3, turbulent transport becomes dominant in determining heat transport. The ion temperature rises from 3.8 keV to 6.1 keV in the core as the rotation velocity increases from 103 rad/s to 105 rad/s. Despite the ion temperature rising, the rotation velocity does not obviously affect electron temperature or density. Additionally, it is noteworthy that the variation in rotation velocity does not significantly affect the global confinement of plasma in scenarios with low density or with high density.

Influence of lower hybrid wave injection on peeling-ballooning modes

The high-confinement mode (H-mode) significantly enhances the energy and particle confinement in fusion plasma compared with the low-confinement mode (L-mode), and it is the basic operation scenario for ITER and CFETR. Edge localized mode (ELM) often appears in H-mode, helping to expel impurities to maintain a longer stable state. However, the particle burst and energy burst from ELM eruptions can severely damage the first wall of fusion device, so, it is necessary to control the ELM. Experiments on EAST tokamak and HL-2A tokamak have been conducted with ELM mitigation by lower hybrid wave (LHW), confirming the effect of LHW on ELMs, but the physical mechanism of ELM mitigation by LHW is still not fully understood. In this paper, the influences of LHW injection on the linear and nonlinear characteristics of peeling-ballooning mode (P-B mode) are investigated in the edge pedestal region of H-mode plasma in tokamak by using the BOUT++ code. The simulations take into consideration both the conventional main plasma current driven by LHW and the three-dimensional perturbed magnetic field generated by the scrape-off layer helical current filament (HCF) on the P-B mode. The linear results show that the core plasma current driven by LHW moves the linear toroidal mode spectrum towards higher mode numbers and lower growth rates by reducing the normalized pressure gradient and magnetic shear of the equilibrium. Nonlinear simulations indicate that due to the broadening of the linear mode spectrum, the core current driven by LHW can reduce the pedestal energy loss caused by ELM through globally suppressing different toroidal modes of the P-B mode, and the three-dimensional perturbed magnetic field generated by LHW-driven HCF can reduce the energy loss caused by ELMs through promoting the growth of modes other than the main mode and enhancing the coupling between different modes. It is found in the study that the P-B mode promoted by the three-dimensional perturbed magnetic field generated by HCF has a mode number threshold, and when the dominant mode of the P-B mode is far from the mode number threshold driven by the three-dimensional perturbed magnetic field, the energy loss due to ELMs is more significantly reduced. These results contribute to a more in-depth understanding of the physical mechanism in ELM control experiment by LHW.

Disruption avoidance and investigation of the H-Mode density limit in ASDEX Upgrade

In recent years a strong effort has been made to investigate disruption avoidance schemes in order to aid the development of integrated operational scenarios for ITER. Within the EUROfusion programme the disruptive H-mode density limit (HDL) has been studied on the WPTE (Work Package Tokamak Exploitation) devices ASDEX Upgrade, TCV and JET. Advanced real-time control coupled with improved real-time diagnostics has enabled the routine disruption avoidance of the HDL. This allowed the systematic study of the influence of various plasma parameters on the onset and behavior of the HDL in regimes not easily accessible otherwise. The upper triangularity δtop is found to have a significant influence on the x-point radiator (XPR), which plays a major role for the evolution of the disruptive HDL. At high δtop the gas flow rate at which the onset of the XPR occurs is strongly reduced compared to low δtop . The reduction of δtop has proven to be an effective actuator for the HDL disruption avoidance on ASDEX Upgrade for highly shaped scenarios ( δtop>0.25 ). It is observed that the occurrence of the XPR and the H–L transition at the density limit are two separate events, the order of which depends on the applied auxiliary heating power. At sufficiently high heating power the XPR occurs before the H–L transition. Impurity seeding, used for divertor detachment, influences the onset and the dynamics of the XPR and the behavior of the HDL. The stable existence of the XPR, which is thought to be a requirement for detachment control in future devices, has also been observed without impurity seeding. The implementation of a robust and sustainable operational scenario, e.g. for ITER, requires the combination of continuous control and exception handling. For each disruption path the appropriate observers and actuators have to be validated in present devices. Automation of the dynamic pulse schedule has proven successful to scan the operational space of the HDL without disruption. Applying such a technique to ITER could reduce the machine risk induced by disruptions during commissioning. The methodology to develop physics-based observers, which indicate the entry into a disruption path well in time, and applying the appropriate action before the discharge becomes unstable has proven successful.

Iterative framework based on multi-task learning for service recommendation

In recent years, service-oriented computing technology has developed rapidly, which, however, has increased the burden of selection for software developers when developing service-based systems. To solve this problem, people have proposed various methods to recommend services which are composed into an application. Nevertheless, most of the existing service recommendation methods cannot serve iterative scenarios, i.e., multiple request–response recommending rounds, which frequently occur in real application development. Moreover, they usually fail to utilize full features such as user requirements and service categories, leading to poor performance of service recommendation. To solve the above problems, we propose an iterative framework for service recommendation through multi-model fusion and multi-task learning, called ISRMM. More specifically, we design two models to capture the preferences of applications towards services, through the perspectives of user requirements and history interaction respectively. The output features of the above models are further fused to predict the next service that will be recommended. In addition, we add a tag judgment task to make our framework capable of multi-task learning, through which, the training signal information implied can be used as an inductive bias to improve service recommendation capabilities. Extensive experiments on real datasets show that ISRMM outperforms several state-of-the-art service recommendation methods in iterative service recommendation scenarios.

Experiments in high-performance JET plasmas in preparation of second harmonic ICRF heating of tritium in ITER

The reference ion cyclotron resonance frequency (ICRF) heating schemes for ITER deuterium–tritium (D-T) plasmas at the full magnetic field of 5.3 T are second harmonic heating of T and 3He minority heating. The wave-particle resonance location for these schemes coincide and are central at a wave frequency of 53 MHz at 5.3 T. Experiments have been carried out in the second major D-T campaign (DTE2) at JET, and in its prior D campaigns, to integrate these ICRF scenarios in JET high-performance plasmas and to compare their performance with the commonly used hydrogen (H) minority heating. In 50:50 D:T plasmas, up to 35% and 5% larger fusion power and diamagnetic energy content, respectively, were obtained with second harmonic heating of T as compared to H minority heating at comparable total input powers and gas injection rates. The core ion temperature was up to 30% and 20% higher with second harmonic T and 3He minority heating, respectively, with respect to H minority heating. These are favourable results for the use of these scenarios in ITER and future fusion reactors. According to modelling, adding ICRF heating to neutral beam injection using D and T beams resulted in a 10%–20% increase of on-axis bulk ion heating in the D-T plasmas due to its localisation in the plasma core. Central power deposition was confirmed with the break-in-slope and fast Fourier transform analysis of ion and electron temperature in response to ICRF modulation. The tail temperature of fast ICRF-accelerated tritons, their enhancement of the fusion yield and time behaviour as measured by an upgraded magnetic proton recoil spectrometer and neutral particle analyser were found in agreement with theoretical predictions. No losses of ICRF-accelerated ions were observed by fast ion detectors, which was as expected given the high plasma density of n e ≈ 7–8 × 1019 m−3 in the main heating phase that limited the formation of ICRF-accelerated fast ion tails. 3He was introduced in the machine by 3He gas injection, and the 3He concentration was measured by a high-resolution optical penning gauge in the sub-divertor region. The DTE2 experiments with 3He minority heating were carried with a low 3He concentration in the range of 2%–4% given the fact that the highest neutron rates with 3He minority heating in D plasmas were obtained at low 3He concentrations of ∼2%, which also coincided with the highest plasma diamagnetic energy content. In addition to 3He introduced by 3He gas injection, an intrinsic concentration of 3He of the order of 0.2%–0.4% was measured in D-T plasmas before 3He was introduced in the device, which is attributed to the radioactive decay of tritium to 3He. According to modelling, even such low intrinsic concentrations of 3He lead to significant changes in ICRF power partitioning during second harmonic heating of T due to absorption of up to 30% of the wave power by 3He.

An IMAS-integrated workflow for energetic particle stability

The confinement of energetic particles (EPs) generated by fusion reactions and external heating methods is crucial for the performance of future fusion devices. However, EP transport can occur due to their interaction with electromagnetic perturbations, affecting heating efficiency and overall performance. Robust reduced models are needed to analyze stability and transport. This paper presents an automated IMAS-based workflow for analyzing the time-dependent stability of EP-driven modes, focusing on the linear properties of Toroidal Alfvén Eigenmodes (TAEs) in general tokamak geometry. The workflow utilizes efficient computational methods and reduced models to deliver fast and reproducible results. A demonstration of the workflow’s effectiveness was performed, identifying key linear properties of TAEs in various simulated ITER scenarios. This approach represents a critical step toward developing tools for analyzing EP transport and optimizing the performance of future fusion reactors.

AC Losses Calculation in the ITER TF Magnets During a Scenario

While the ITER Magnet system is being assembled at Cadarache (France), models are being developed on the basis of the conductors test results. The purpose is to prepare the performances extrapolation to magnet commissioning and operation. In particular, the AC losses generated in the conductors during a plasma scenario should be estimated in order to predict potential operating margins reductions and to prevent quenches. In particular, there is the case of AC losses in the conductors of the TF inner leg, where the field variations are both transverse and parallel to the axis of the conductor, during a PF/CS current variation. This paper proposes firstly to review the TF conductors results, extracted from the qualification phase, defining a reduced set of parameters for AC-losses calculation. Then, we propose expressions for calculating losses caused by a parallel field variation. Finally, we analyze a standard ITER scenario, and compute the generated losses, to later be introduced as inputs for thermohydraulic calculations.

A comparison of the influence of plasmoid-drift mechanisms on plasma fuelling by cryogenic pellets in ITER and Wendelstein 7-X

Pellet injection is the most promising technique to achieve efficient plasma core fuelling, key for attaining stationary scenarios in large magnetic confinement fusion devices. In this paper, the injection of pellets with different volumes and speeds into standard plasma scenarios in ITER (tokamak) and Wendelstein 7-X (stellarator) is studied by modeling the pellet ablation and particle deposition, focusing on the evaluation of the expected differences in pellet plasmoid drifts in tokamaks and stellarators. Since the efficiency of the damping-drift mechanisms is predicted to depend on the magnetic configuration, device-specific characteristics are expected for the temporal evolution of the plasmoid drift acceleration. For instance, plasmoid-internal Pfirsch–Schlüter currents dominate the drift damping process for stellarators, while plasmoid-external currents are more relevant for tokamaks. Also, relatively larger drifts are in principle expected for W7-X due to higher field gradients in relation to machine dimensions. However, shorter plasmoid-internal charge reconnection lengths result in the drift damping due to internal Pfirsch–Schlüter currents being more effective than in a tokamak. Therefore, the average relative drift displacement during the whole plasmoid homogenization may a priori be comparable in both magnetic configurations. Moreover, High Field Side (HFS) injection is expected to be highly advantageous to maximize pellet particle deposition in ITER, whereas it may only be beneficial in medium to high β environments in W7-X. Finally, there may be means for the optimization of pellet injection configurations in both ITER and W7-X for the considered plasma scenarios despite the sizeable differences in the relative importance of the mechanisms of plasmoid drift acceleration and deceleration in play.

Toward a Super-COP? Timing, Temporality, and Rethinking World Climate Governance

Toward a Super-COP? Timing, Temporality, and Rethinking World Climate Governance

Parametric Decay Instabilities during Electron Cyclotron Resonance Heating of Fusion Plasmas, Problems and Possibilities

We review parametric decay instabilities (PDIs) expected in connection with electron cyclotron resonance heating (ECRH) of magnetically confined fusion plasmas, with a specific focus on conditions relevant for the ITER tokamak. PDIs involving upper hybrid (UH) waves are likely to occur in O-mode ECRH scenarios at ITER if electron density profiles allowing trapping of UH waves near the ECRH frequency are present. Such PDIs may occur near the plasma center in ITER full-field scenarios heated by 170 GHz O-mode ECRH and on the high-field side of half-field ITER plasmas heated by 110 GHz or 104 GHz O-mode ECRH. Additionally, 110 GHz O-mode ECRH of half-field ITER scenarios may have low ECRH absorption, due to the electron cyclotron resonance being located on the high-field side of the main plasma. This potentially allows PDIs driven by a significant amount of ECRH radiation reaching the UH resonance in X-mode to occur, as X-mode radiation can be generated by reflection of unabsorbed O-mode radiation from the high-field side wall. The occurrence of PDIs during ECRH may damage microwave diagnostics, such as the electron cyclotron emission and low-field side reflectometer systems at ITER, as well as complicate the calculation of heating and current drive characteristics. However, if PDIs are induced in a controlled manner, they may provide novel diagnostic tools and allow the generation of a moderate fast ion population in plasmas heated only by ECRH.

Data-driven robust optimization using deep neural networks

Robust optimization has been established as a leading methodology to approach decision problems under uncertainty. To derive a robust optimization model, a central ingredient is to identify a suitable model for uncertainty, which is called the uncertainty set. An ongoing challenge in the recent literature is to derive uncertainty sets from given historical data that result in solutions that are robust regarding future scenarios. In this paper we use an unsupervised deep learning method to learn and extract hidden structures and anomalies from data, leading to non-convex uncertainty sets and better robust solutions. We prove that most of the classical uncertainty classes are special cases of our derived sets and that optimizing over them is strongly NP-hard. Nevertheless, we show that the trained neural networks can be integrated into a robust optimization model by formulating the adversarial problem as a convex quadratic mixed-integer program. This allows us to derive robust solutions through an iterative scenario generation process. In our computational experiments, we compare this approach to a similar approach using kernel-based support vector clustering and to other benchmark methods. We find that uncertainty sets derived by the unsupervised deep learning method find a better description of data and lead to robust solutions that often outperform the comparison methods both with respect to objective value and feasibility.

A Systemic Approach for Natural Language Scenario Elicitation of Security Requirements

Security analysts rely on scenarios to assess vulnerabilities, project attacks, and decide on security requirements that mitigate the threat. However, eliciting natural language scenarios from stakeholders can be an ad-hoc process and subject to ambiguity and incompleteness. In this article, we examine systematic scenario elicitation by introducing a method based on user stories that uses a simplified process model of iterative scenario refinement. The process consists of three steps: 1) eliciting an interaction statement that describes a critical action performed by a user or system process; 2) eliciting one or more descriptive statements about a technology that enables the interaction; and 3) refinement of the technology into technical variants that correspond to design alternatives. We empirically evaluated our method by implementing our prototype in a user study that collects 30 security scenarios from participants. Based on our analysis, our proposed method is shown effective. Participants had a 100 percent task completion rate with 57 percent of participants achieving complete task-success, and the remaining 43 percent of participants achieving partial task-success. We also show the effect of security domain knowledge, and the benefit of using structure when collecting security requirements in natural language text. Finally, we present lessons learned and future research directions.

Analysis of the highest energy confinement in tokamaks based on thermodynamic approach

We analyze the highest energy confinement in tokamak plasmas based on thermodynamic approach (plasma self-organization). The energy transport coefficients in the saturated confinement regimes are calculated from experiments in the T-10 tokamak. Using these coefficients, we estimate the maximal energy confinement for JET, ASDEX Upgrade, JT-60U, DIII-D and KSTAR tokamaks. Calculated energy confinement is in a good agreement with measured ones. Obtained results allow us to predict the maximal energy confinement in newly constructed machines up to a fusion reactor. The energy confinement for two basic scenarios for ITER is accessed.