roles of the major us toroidal magnetic fusion facilities a summary of the 2005 fesac facilities...

Roles of the Major US ToroidalMagnetic Fusion Facilities

A Summary of the 2005 FESAC Facilities Panel Report

Presented by: Jill Dahlburg Naval Research Laboratory, Washington, DC

with: Steven Allen (LLNL), Riccardo Betti (UR), Stephen Knowlton (Auburn), Rajesh Maingi (ORNL), Gerald Navratil (Columbia), Steve Sabbagh (Columbia), John Sheffield (U.TN), James Van Dam (UT-Austin), Dennis Whyte (U.Wisc)

# 1 of 34: October 12, 2005

5 April 2005 FESAC Charge

Dear Professor Hazeltine:

Funding U.S. participation in the ITER project will place considerable pressure on the Office of Sciencebudget in future years. I am committed to a strong base fusion research program during ITER constructionbut want to be sure that the program is cost-effective. In particular, I do not believe that we should continuerunning existing facilities just because we can do more research on them but because we can do unique andimportant research on them. Accordingly, I would like you to a ssess characteristics of the three major U.S.toroidal magnetic fusion facilities in the context of the international fusion programs and determine whatcontributions they can make to fusion science and the fundamental vitality of the U.S. Fusion Program duringthe next five years.

I would like you to consider the following issues in your assessment:

1. What are the unique and complementary characteristics of each of the major U.S. fusion facilities? 2. How do the characteristics of each of the three U.S. fusion facilities make the U.S. toroidal research program unique as a whole in the international program? 3. How well do we cooperate with the international community in coordinating research on our major facilities and how have we exploited the special features of U.S. facilities in contributing to international fusion research, in general, and to the ITER design specifically? 4. How do these three facilities contribute to fusion science and the vitality of the U.S. Fusion program? What research opportunities would be lost by shutting down one of the major facilities?

Sincerely, Raymond L. Orbach, Director, Office of Science

OverviewOverview# 2 of 34: October 12, 2005

The report addresses the three major US toroidal fusion facilities.

The three major toroidal fusion research facilities in the US have diverse and complementary characteristics, which were

developed on the basis of evolving US innovation in fusion energy sciences.

Taken together these three facilities provide the US with a very effective presence in the world program of fusion research.

Their success has enabled the US to have substantial impact on the direction and progress of the field, including leadership in

understanding fundamental transport processes in magnetically confined plasmas, and continuing optimization of the magnetic

configuration for confinement of high pressure plasmas.


Unique and complementary characteristics

The characteristics that make each of these toroidal facilities unique in research capability stem from their initial research motivations:

C-Mod to understand plasma behavior at very high magnetic field with the plasma pressure and field appropriate for sustaining a burning plasma for energy production, but at a smaller size than a self-heated plasma.

DIII-D to understand and improve plasma confinement and stability as a function of plasma shape and magnetic field distribution in a collisionless plasma.

NSTX (the most recently commissioned, in 1999) to apply advances in understanding of magnetic field configuration to optimize both plasma stability and confinement in a proof-of-principle, low aspect ratio toroidal experiment (the spherical torus).


One standard for judging the progress in fusion ..

.. is the “triple product” of plasma density, temperature, and confinement time, which may be rephrased as the product of: plasma (plasma pressure / field pressure); the square of the magnetic field; and, the confinement time. In this rephrased form, the triple product delineates the three approaches to fusion embodied in the three facilities:

C-Mod is the highest magnetic field, diverted tokamak in the world;

DIII-D pursues the direction of high confinement through the development of advanced tokamak operational regimes with long confinement times;

NSTX pursues the direction of very high , resulting from its extreme toroidicity (i.e., low aspect ratio).


Today, each facility is a leading element of the world program in magnetic fusion research.

C-Mod is one of two tokamaks tied with world's highest tokamak magnetic field and is the only facility capable of studying plasma/ wall interactions and radio frequency heating and current drive in ITER-like geometry with a divertor and magnetic field and plasma pressure characteristic of a burning plasma.

DIII-D, with its unparalleled plasma transport diagnostics and world-leading capability for plasma shaping and control of major instabilities limiting high pressure plasmas, has established itself as a center for developing long-pulse, high performance advanced tokamak operation.

NSTX, the world’s most capable spherical torus, explores high- plasma stability and confinement at extreme toroidicity (low aspect ratio) and is the major US experiment for concept innovation in magnetic confinement now in operation.

While these three facilities are clearly distinct, they also have a degree of commonality that makes them highly effective as a group.


The panel considered the charge questions [#1, #2 & #4i] from the perspective of ...

.. the most important frontier areas in fusion energy sciences research, recently described in detail in the April 2005 FESAC Program Priorities Report:

Macroscopic plasma physics: Understand the role of magnetic structure on plasma confinement and the limits to plasma pressure in sustained magnetic configurations;

Multi-scale transport physics: Understand and control the physical processes that govern the confinement of heat, momentum, and particles in plasmas;

Plasma boundary interfaces: Learn to control the interface between the 100-million- degree-C plasma and its room temperature surroundings;

Waves and energetic particles: Learn to use waves and energetic particles to sustain and control high temperature plasmas;

Fusion engineering science: Understand the fundamental properties of materials, and the engineering science of the harsh fusion environment.




Macroscopic plasma physics: Understand the role of magnetic structure on plasma confinement and the limits to plasma pressure in sustained magnetic configurations.





OverviewOverview

Macroscopic Plasma Physics

Macroscopic plasma physics seeks to determine how to confine and sustain maximum plasma

pressure efficiently in a magnetic field configuration.

This is extremely important since fusion energy production in a burning plasma (such as ITER)

increases with the square of the plasma pressure.

Macroscopic Plasma PhysicsMacroscopic Plasma Physics# 8 of 34: October 12, 2005


Figure 2.1-1: DIII-D can produce a wide range of plasma shapes and can match the shapes of most machines, including the ITER design.

1.2 < κ < 2.40 < δ < 1squarenessindentation2.6 < < 3.5A

C-Mod: operates at, or above toroidal field of ITER (up to 8T)

DIII-D: most flexible shaping; can produce a wide range of shapes, including matching the shapes of most machines

NSTX: low aspect ratio allows very high elongation, enabling very high


Plasma shape can be altered to increase plasma pressure. All 3 facilities can match the ITER cross-sectional shape:


Instabilities can cause rapid loss of plasma pressure and current.

- DIII-D, NSTX produce, and are diagnosed to study RWM, NTM, ELMs- C-Mod, DIII-D can study avoidance, mitigation of disruptions

(a) (b) (c)

Figure 2.1-2: Resistive wall mode (RWM) in DIII-D and NSTX: (a) DIII-D plasma stabilized by rotation; (b) unstable n = 1 RWM in DIII-D; and (c) RWM with n = 1-3 components in NSTX (surface distortions shown 10x exaggerated).

114147 NSTXDIII–DDIII–D


Strength of the three facilities in combination enables addressing

ITER plasma can be quickly lost, in a “disruption”, resulting in:– Large forces on the tokamak mechanical structure

– Large heat loads to the wall

The U.S. has been a leader in disruption experiments and modeling:– Detect a disruption, and– Apply a large puff of gas to safely extinguish the plasma– DIII-D did the first experiments– Work expanded by C-Mod to high absolute pressure plasma– Both experiments and modeling, along with JET, are being used to

develop ITER scenarios– NSTX shows moderate resilience against disruptions: science

understanding

machine damage due to sudden disruptions:





Multi-scale transport physics: Understand and control the physical processes that govern the confinement of heat, momentum, and particles in plasmas.




OverviewOverview

Multi-scale Transport Physics

The U.S. leads the international research effort to determine the instabilities responsible for turbulent transport.

The combination of the three machines improves ability to separate out the variables that control energy confinement through cross-machine studies.

– Combination of facilities also makes it possible to perform cross-machine identity and similarity experiments.

The U.S. facilities have diagnostics covering a wide range of inverse wavelengths k, from ion to electron scales.

– Although ion transport is reasonably well understood, electron transport is not.

Multi-scale Transport PhysicsMulti-scale Transport Physics# 12 of 34: October 12, 2005

Capabilities to Study Transport

C-Mod• Studies at high field

and power density (for same N and *), with Ti = Te

• Particle/momentum-free heating

• Novel diagnostics

DIII-D• Integrated control of

shape, profiles, and rotation

• Comprehensive diagnostics

• Advanced gyrokinetic simulation code

NSTX• Studies at high beta

(EM effects) and low R/a, with rotation

• Novel diagnostics



Transport basis for ITER operating scenarios– U.S. effort focused around Transport Task Force (TTF) and ITPA

Strong diagnostic effort on all three machines– C-Mod studies confinement at high ne with Te~Ti and novel PCI diagnostic

– DIII-D has comprehensive diagnostic set and has developed the most comprehensive transport code for experiment-theory comparisons

– NSTX examines how transport scales with aspect ratio at high , commissioning fluctuation diagnostics

Data from all three machines are important for accurate determination of relevant scaling variables

– ITPA confinement scaling database

The U.S. Transport program has improved - and will continue to improve - the reliability of transport predictions for ITER

the transport basis for ITER:






Plasma boundary interfaces: Learn to control the interface between the 100-million- degree-C plasma and its room temperature surroundings.



OverviewOverview

Plasma Boundary Interface

The FESAC Priorities Panel report topical science question:“How can a 100-million-degree burning plasma be interfaced to its room temperature surroundings?”

Critical to operational and scientific success of ITER, which will have unprecedented pulse length, energy density, power loads and tritium fuel throughput

Timely contributions from US facilities in next 5 years on– Plasma-facing materials– Pedestal physics– Edge localized modes– Tritium retention & recovery– Boundary layer particle transport

Plasma Boundary InterfacesPlasma Boundary Interfaces# 15 of 34: October 12, 2005

US boundary research makes vital contributions to ITER

Divertor magnetic topology is used on all three US facilities and ITER

C-Mod & DIII-D developed and diagnosed divertor detachment to reduce target peak power loads to acceptable levels in ITER

ITER adopted vertical target geometry of C-Mod to facilitate detachment

DIII-D demonstrated density control with pumping of highly shaped plasmas for AT studies as planned for ITER

Effect of radiation opacity on detachment in high density C-Mod divertor for ITER



tritium inventory and choice of wall materials:

Current ITER has Be walls, with tungsten carbon in the divertor

Carbon can handle high transient heat loads, – But is co-deposited with tritium

– Can lead to large in-vessel tritium inventory

US machines are attacking the problem in different ways, complement JET & ASDEX-U.

Carbon: DIII-D and NSTX (with lithium)– Further characterize carbon erosion and redeposition

– Study tritium removal techniques

Metal: C-Mod– Molybdenum and Tungsten do not lead to tritium accumulation– Can melt with disruptions and radiate strongly in the core







Waves and energetic particles: Learn to use waves and energetic particles to sustain and control high temperature plasmas.


OverviewOverview

Waves and Energetic Particle Research are Integral Component of US Fusion Program

High power, externally launched radio-frequency waves (30 MHz

to 110 GHz) heat plasmas and drive non-inductive toroidal current– Only method for heating core of large burning plasma

– Precision methods to control heating, current, and possibly plasma flow; crucial to innovations of “advanced tokamak” scenarios for ITER and beyond

Non-thermal particle distributions, e.g., fusion-produced -particles in ITER, may destabilize Alfven wave modes of the plasma

– Possibility of reduced confinement and enhanced macroscopic instability in burning plasmas

Waves & Energetic ParticlesWaves & Energetic Particles# 18 of 34: October 12, 2005

Variety of Wave Heating Schemes in Distinct Plasma Conditions in 3 US facilities

Missions: plasma heating, local current drive, instability control

DIII-D: Moderate field, moderate density advanced tokamakElectron cyclotron waves for current drive (broader current profile, control of tearing

instabilities); selected for ITER

Fast wave (ion cyclotron frequency range) for current replacement in core

NSTX: Spherical tori address non-inductive current profile control in “overdense” plasmas for long-pulse operation

Electron Bernstein wave current drive to be initiated for current drive.

Fast wave for core current drive.

Coaxial Helicity Injection for non-inductive start-up.

C-Mod: Entirely radio-frequency wave-heated, high-density, high-field plasmasIon cyclotron minority heating delivers flexible, controlled heating; selected for ITER.

Ion cyclotron mode conversion to supply local current drive and plasma flow

Lower hybrid current drive was recently implemented to access advanced tokamak schemes; this is a reserve option for ITER.


wave and energetic particle physics:

US facilities pursue non-duplicative methods of wave heating

and current drive appropriate to diverse missions ofMHD instability suppression

Control of the radial profile of plasma current

Localized plasma heating

in significantly different parameter ranges of the experiments.

Studies performed in next five years will be important for ITER, and crucial for advanced tokamak/ innovative concept development.

US facilities contribute significantly to vitally important understanding of threat of Alfven eigenmodes to burning plasma confinement.










OverviewOverview

e.g., plasma facing materials

An issue for ITER which plans Beryllium main walls, Carbon fiber composites and Tungsten brushes for the divertor

C-Mod has all-metal, molybdenum walls, and is testing tungsten brushes built by Sandia – see figure

DIII-D uses all carbon and may test hydrogen recovery with oxygen baking

NSTX uses carbon and is testing lithium for pumping and as a divertor target

C-Mod and NSTX target plates have ITER-level divertor power density

Tungsten “brush” tile built by Sandia is being

tested in C-Mod

Fusion Engineering Science

Fusion Engineering ScienceFusion Engineering Science# 21 of 34: October 12, 2005

Contributions to & Cooperation withthe International Community

Research contributions made by the US facilities to: — International program in burning plasma research— Future plans with emphasis on ITER

FESAC Priorities Panel Themes:– “Create a Star on earth”– “Develop the science and technology to realize fusion energy”

Coordinating framework is International Tokamak Physics Activity– Over 50 US scientists, overall head is a US scientist– ITPA addresses comprehensive set of science issues, including:

• Confinement database and modeling• Transport• Pedestal and Edge Physics• Divertor and Scrape-Off Layer• MHD control and disruptions• Steady-state operation• Diagnostics

InternationalInternational# 22 of 34: October 12, 2005

Science on US facilities will prepare us for effective participation in ITER

Increase confidence in current ITER design– e.g., Choice of wall materials

Provide information for design decisions not yet finalized– e.g., Details of heating systems

Suggest possible improvements to baseline design– e.g., Magnetic control coils for stabilizing MHD modes

Develop new measurement techniques, diagnostics, control systems– e.g., Sensing and mitigation of plasma disruptions

Enhance theory and integrated modeling– e.g., Design of experiments for ITER

InternationalInternational# 23 of 34: October 12, 2005

Integration of the fundamental science and technology issues discussed in the FESAC Facilities Panel Report

Thirty-five Years of Research

US toroidal magnetic fusion research has developed from a collection of independent institutions and research activities into a cohesive scientific research program that has as its basis a complementary trio of collaborative national experimental research facilities. A consequence of this melding is that researchers in the US agree that the ‘knowledge base is now in hand to produce a burning plasma – a plasma whose high temperature is sustained predominantly by energy from alpha particles produced by fusion reactions.’

This confidence results from a number of US achievements in the last decade, including demonstrated control of tokamak major disruptions, the enhancement of ion confinement through sheared flows and transport barriers, the first-ever studies of burning plasma behavior, and the sustained ability to operate high performance plasmas in the MHD-stable advanced tokamak mode, with high self-driven current, good particle and energy confinement, and a plasma edge that enables particle and power handling.

ITER will benefit significantly from these accomplishments.

What would be lost, and RecommendationWhat would be lost, and Recommendation# 24 of 34: October 12, 2005

Alcator C-Mod

Alcator C-Mod is distinguished by the following salient characteristics: (1) it operates at higher magnetic fields than any other existing divertor tokamak, and over a range that spans the ITER magnetic field value; (2) it has all-metallic (high atomic number) tungsten and molybdenum wall armor and divertor plates; and (3) its non-inductive heating and current drive are supplied entirely by flexible radio frequency techniques. C-Mod is also distinguished by its location at a major research university and, as such, hosts the largest number of graduate and undergraduate students.


What Research Opportunities Would Be Lost

The loss of C-Mod would:

Eliminate studies of tritium fuel retention and power flow to the wall and divertor in ITER-relevant edge conditions, which have potentially serious consequences for ITER.

Eliminate ITER-relevant tests of thermal, particle and momentum transport with radio frequency heating and current-drive techniques in plasmas with the ITER-like characteristics of ions and electrons coupled at the same temperature through collisions, no core external momentum-drive, and high-pressure edge conditions.

Compromise development of ion cyclotron and lower hybrid RF capabilities for controlled plasma heating and edge current drive essential for advanced tokamak scenarios in ITER and future burning plasma facilities.


DIII-D

DIII-D is the best equipped – in terms of diagnostics, heating and control systems – and most flexible tokamak –in terms of shaping capability – in the world. It has an outstanding record of contributing to plasma science and technology, and to the development of advanced scenarios for tokamak operation that offer the possibility for ITER to achieve its goals at reduced plasma current. Its capabilities have been enhanced, and this excellent program is expected to continue producing centrally important information for the advancement of science and the magnetic fusion program.



The loss of DIII-D would:

Eliminate a world-class program that contributes to understanding turbulent transport, fast ion instabilities, pedestal and divertor physics, and mode stabilization.

Greatly reduce US leadership in the world-wide development of a high , high bootstrap current, "steady-state" plasma and of hybrid operating scenarios for ITER.

Cede the US position of strength in electron cyclotron current drive for current profile and MHD stability control.


NSTX

NSTX has exceptional performance and diagnostic capability among spherical tori, with highly flexible shaping capability and the ability to access ultra-high plasma values. The facility provides an operational regime for key science theories to be tested and confirmed, providing understanding for tokamak fusion devices in general.



The loss of NSTX would:

Eliminate not only the US leadership position in the world for device capability and research on spherical tokamaks, but also eliminates the U.S. ability to guide and contribute to spherical torus research at the “proof-of-principle” level.

Eliminate numerous important experiments such as Bernstein wave current drive, plasma start-up in a proof-of-principle scale experiment, ultra-high operation, and fast ion measurements.

Eliminate US leadership in high spherical tokamak research, and the U.S. independence to pave the path to the future strategic option of constructing a Component Test Facility based on the spherical torus.


US Position Internationally

In a February 2005 report titled ‘The Knowledge Economy: is the United States losing its competitive edge?,’ the Task Force on the Future of American Innovation “developed a set of benchmarks to assess the international standing of the US in science and technology. These benchmarks in education, the science and engineering workforce, scientific knowledge, innovation, investment and high-tech economic output reveal troubling trends across the research and development spectrum. The US still leads the world in research and discovery, but our advantage is rapidly eroding, and our global competitors may soon overtake us.”

In the area of magnetic fusion energy research, the Facility Panel found that the US at present holds a position of international strength and leadership. The next major magnetic fusion research facility will be the offshore plasma experiment, ITER. Fostered by the recent decision over its ITER site, interest in ITER is rapidly growing in the international research community. The loss of any of the three major U.S. toroidal fusion facilities would fundamentally jeopardize the ability of U.S. researchers to perform relevant fusion research, and thus would undermine the current US position of international excellence.


What would be lost?

What would be lost, and RecommendationWhat would be lost, and Recommendation

It is important for the US to to be viewed as a major player in the world program in order to effect the science on and reap full benefit from ITER:

only a folio of good research results buys a seat at that sort of decision table.

Report Recommendation

The three major US magnetic fusion facilities represent a massive investment of talent, intellect, and finances in tackling the key issues of toroidal confinement.

Each has made seminal contributions to the development of toroidal confinement and to the fundamental S&T that undergird it. The wealth of discoveries and the generation of knowledge made possible by the three coordinated US facilities has enabled the US to be an effective presence among the larger foreign programs involved in ITER.

Premature closure of one of these major facilities would seriously compromise the effectiveness of the US fusion program internationally

and also the US ability to advocate future proposals for advanced performance scenarios that could lead to a more economically

competitive high-power-density fusion system.


Report Recommendation

The Panel’s recommendation is that the three major United States toroidal magnetic fusion facilities continue

operation to conduct important unique and complementary research in support

of fusion energy sciences and ITER.

RecommendationRecommendation# 33 of 34: October 12, 2005

D I S C U S S I O N

* Photo/Image provided courtesy of the Naval Research Laboratory.

NRL spectroheliograph image of the sun, taken aboard Skylab in 1973, using the

extreme ultraviolet radiation from ionized helium, 304 Angstom wavelength.

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# 34 of 34: October 12, 2005

roles of the major us toroidal magnetic fusion facilities a summary of the 2005 fesac facilities...

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