B. C. Stratton

Bio: B. C. Stratton is an academic researcher from Princeton Plasma Physics Laboratory. The author has contributed to research in topics: Tokamak Fusion Test Reactor & Tokamak. The author has an hindex of 16, co-authored 59 publications receiving 1308 citations.

Enhancement of Tokamak Fusion Test Reactor performance by lithium conditioning

Abstract: Wall conditioning in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] by injection of lithium pellets into the plasma has resulted in large improvements in deuterium–tritium fusion power production (up to 10.7 MW), the Lawson triple product (up to 1021 m−3 s keV), and energy confinement time (up to 330 ms). The maximum plasma current for access to high‐performance supershots has been increased from 1.9 to 2.7 MA, leading to stable operation at plasma stored energy values greater than 5 MJ. The amount of lithium on the limiter and the effectiveness of its action are maximized through (1) distributing the Li over the limiter surface by injection of four Li pellets into Ohmic plasmas of increasing major and minor radius, and (2) injection of four Li pellets into the Ohmic phase of supershot discharges before neutral‐beam heating is begun.

Simulations of deuterium-tritium experiments in TFTR

Abstract: A transport code (TRANSP) is used to simulate future deuterium-tritium (DT) experiments in TFTR. The simulations are derived from 14 TFTR DD discharges, and the modelling of one supershot is discussed in detail to indicate the degree of accuracy of the TRANSP modelling. Fusion energy yields and alpha particle parameters are calculated, including profiles of the alpha slowing down time, the alpha average energy, and the Alfven speed and frequency. Two types of simulation are discussed. The main emphasis is on the DT equivalent, where an equal mix of D and T is substituted for the D in the initial target plasma, and for the D0 in the neutral beam injection, but the other measured beam and plasma parameters are unchanged. This simulation does not assume that alpha heating will enhance the plasma parameters or that confinement will increase with the addition of tritium. The maximum relative fusion yield calculated for these simulations is QDT ~ 0.3, and the maximum alpha contribution to the central toroidal β is βα(0) ~ 0.5%. The stability of toroidicity induced Alfven eigenmodes (TAE) and kinetic ballooning modes (KBM) is discussed. The TAE mode is predicted to become unstable for some of the simulations, particularly after the termination of neutral beam injection. In the second type of simulation, empirical supershot scaling relations are used to project the performance at the maximum expected beam power. The MHD stability of the simulations is discussed

Fusion power production from TFTR plasmas fueled with deuterium and tritium

Abstract: Peak fusion power production of 6.2 ± 0.4 MW has been achieved in TFTR plasmas heated by deuterium and tritium neutral beams at a total power of 29.5 MW. These plasmas have an inferred central fusion alpha particle density of 1.2 x 1017 m ₋3 without the appearance of either disruptive magnetohydrodynamics events or detectable changes in Alfven wave activity. The measured loss rate of energetic alpha particles agreed with the approximately 5% losses expected from alpha particles which are born on unconfined orbits.

Nondimensional transport scaling in the Tokamak Fusion Test Reactor: Is tokamak transport Bohm or gyro‐Bohm?

Abstract: General plasma physics principles state that power flow Q(r) through a magnetic surface in a tokamak should scale as Q(r)= {32π2Rr3Te2c nea/[eB (a2−r2)2]} F(ρ*,β,ν*,r/a,q,s,r/R,...) where the arguments of F are local, nondimensional plasma parameters and nondimensional gradients. This paper reports an experimental determination of how F varies with normalized gyroradius ρ*≡(2TeMi)1/2c/eBa and collisionality ν*≡(R/r)3/2qRνe(me/ 2Te)1/2 for discharges prepared so that other nondimensional parameters remain close to constant. Tokamak Fusion Test Reactor (TFTR) [D. M. Meade et al., in Plasma Physics and Controlled Nuclear Fusion Research, 1990, Proceedings of the 13th International Conference, Washington (International Atomic Energy Agency, Vienna, 1991), Vol. 1, p. 9] L‐mode data show F to be independent of ρ* and numerically small, corresponding to Bohm scaling with a small multiplicative constant. By contrast, most theories predict gyro‐Bohm scaling: F∝ρ*. Bohm scaling implies that the largest scale size f...

Alpha Particle Physics Experiments in the Tokamak Fusion Test Reactor

Abstract: Alpha particle physics experiments were done on TFTR during its DT run from 1993 to 1997. These experiments utilized several new alpha particle diagnostics and hundreds of DT discharges to characterize the alpha particle confinement and wave-particle interactions. In general, the results from the alpha particle diagnostics agreed with the classical single particle confinement model in MHD quiescent discharges. The alpha loss due to toroidal field ripple was identified in some cases, and the low radial diffusivity inferred for high energy alphas was consistent with orbit averaging over small scale turbulence. Finally, the observed alpha particle interactions with sawteeth, toroidal Alfven eigenmodes and ICRF waves were approximately consistent with theoretical modelling. What was learned is reviewed and what remains to be understood is identified.

Development of the indirect‐drive approach to inertial confinement fusion and the target physics basis for ignition and gain

Abstract: Inertial confinement fusion (ICF) is an approach to fusion that relies on the inertia of the fuel mass to provide confinement. To achieve conditions under which inertial confinement is sufficient for efficient thermonuclear burn, a capsule (generally a spherical shell) containing thermonuclear fuel is compressed in an implosion process to conditions of high density and temperature. ICF capsules rely on either electron conduction (direct drive) or x rays (indirect drive) for energy transport to drive an implosion. In direct drive, the laser beams (or charged particle beams) are aimed directly at a target. The laser energy is transferred to electrons by means of inverse bremsstrahlung or a variety of plasma collective processes. In indirect drive, the driver energy (from laser beams or ion beams) is first absorbed in a high‐Z enclosure (a hohlraum), which surrounds the capsule. The material heated by the driver emits x rays, which drive the capsule implosion. For optimally designed targets, 70%–80% of the d...

Plasma{material interactions in current tokamaks and their implications for next step fusion reactors

Abstract: The major increase in discharge duration and plasma energy in a next step DT fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety and performance. Erosion will increase to a scale of several centimetres from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma facing components. Controlling plasma-wall interactions is critical to achieving high performance in present day tokamaks, and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of the important consequences of these phenomena stimulated an internationally co-ordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor project (ITER), and significant progress has been made in better understanding these issues. The paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next step fusion reactors. Two main topical groups of interaction are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation and (ii) tritium retention and removal. The use of modelling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R&D avenues for their resolution are presented.

Chapter 1: Overview and summary

Abstract: The ITER Physics Basis presents and evaluates the physics rules and methodologies for plasma performance projections, which provide the basis for the design of a tokamak burning plasma device whose goal is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. This Chapter summarizes the physics basis for burning plasma projections, which is developed in detail by the ITER Physics Expert Groups in subsequent chapters. To set context, the design guidelines and requirements established in the report of ITER Special Working Group 1 are presented, as are the specifics of the tokamak design developed in the Final Design Report of the ITER Engineering Design Activities, which exemplifies burning tokamak plasma experiments. The behaviour of a tokamak plasma is determined by the interaction of many diverse physics processes, all of which bear on projections for both a burning plasma experiment and an eventual tokamak reactor. Key processes summarized here are energy and particle confinement and the H-mode power threshold; MHD stability, including pressure and density limits, neoclassical islands, error fields, disruptions, sawteeth, and ELMs; power and particle exhaust, involving divertor power dispersal, helium exhaust, fuelling and density control, H-mode edge transition region, erosion of plasma facing components, tritium retention; energetic particle physics; auxiliary power physics; and the physics of plasma diagnostics. Summaries of projection methodologies, together with estimates of their attendant uncertainties, are presented in each of these areas. Since each physics element has its own scaling properties, an integrated experimental demonstration of the balance between the combined processes which obtains in a reactor plasma is inaccessible to contemporary experimental facilities: it requires a reactor scale device. It is argued, moreover, that a burning plasma experiment can be sufficiently flexible to permit operation in a steady state mode, with non-inductive plasma current drive, as well as in a pulsed mode where current is inductively driven. Overall, the ITER Physics Basis can support a range of candidate designs for a tokamak burning plasma facility. For each design, there will remain a significant uncertainty in the projected performance, but the projection methodologies outlined here do suffice to specify the major parameters of such a facility and form the basis for assuring that its phased operation will return sufficient information to design a prototype commercial fusion power reactor, thus fulfilling the goal of the ITER project.

Chapter 2: Plasma confinement and transport

Abstract: The understanding and predictive capability of transport physics and plasma confinement is reviewed from the perspective of achieving reactor-scale burning plasmas in the ITER tokamak, for both core and edge plasma regions. Very considerable progress has been made in understanding, controlling and predicting tokamak transport across a wide variety of plasma conditions and regimes since the publication of the ITER Physics Basis (IPB) document (1999 Nucl. Fusion 39 2137-2664). Major areas of progress considered here follow. (1) Substantial improvement in the physics content, capability and reliability of transport simulation and modelling codes, leading to much increased theory/experiment interaction as these codes are increasingly used to interpret and predict experiment. (2) Remarkable progress has been made in developing and understanding regimes of improved core confinement. Internal transport barriers and other forms of reduced core transport are now routinely obtained in all the leading tokamak devices worldwide. (3) The importance of controlling the H-mode edge pedestal is now generally recognized. Substantial progress has been made in extending high confinement H-mode operation to the Greenwald density, the demonstration of Type I ELM mitigation and control techniques and systematic explanation of Type I ELM stability. Theory-based predictive capability has also shown progress by integrating the plasma and neutral transport with MHD stability. (4) Transport projections to ITER are now made using three complementary approaches: empirical or global scaling, theory-based transport modelling and dimensionless parameter scaling (previously, empirical scaling was the dominant approach). For the ITER base case or the reference scenario of conventional ELMy H-mode operation, all three techniques predict that ITER will have sufficient confinement to meet its design target of Q = 10 operation, within similar uncertainties.

Nonlinear gyrokinetic equations for low-frequency electromagnetic waves in general plasma equilibria

Abstract: A nonlinear gyrokinetic formalism for low‐frequency (less than the cyclotron frequency) microscopic electromagnetic perturbations in general magnetic field configurations is developed. The nonlinear equations thus derived are valid in the strong‐turbulence regime and contain effects due to finite Larmor radius, plasma inhomogeneities, and magnetic field geometries. The specific case of axisymmetric tokamaks is then considered and a model nonlinear equation is derived for electrostatic drift waves. Also, applying the formalism to the shear Alfven wave heating scheme, it is found that nonlinear ion Landau damping of kinetic shear‐Alfven waves is modified, both qualitatively and quantitatively, by the diamagnetic drift effects. In particular, wave energy is found to cascade in wavenumber instead of frequency.