Iter Physics

C Wendell Horton, Jr, Sadruddin Benkadda

• Publisher: WSPC
• Year: 2015
• eBook ISBN: 9789814678681
• Topic: Biological Sciences
• Subtopic: Science General
• Index: Biological Sciences

The promise of a vast and clean source of thermal power drove physics research for over fifty years and has finally come to collimation with the international consortium led by the European Union and Japan, with an agreement from seven countries to build a definitive test of fusion power in ITER. It happened because scientists since the Manhattan project have envisioned controlled nuclear fusion in obtaining energy with no carbon dioxide emissions and no toxic nuclear waste products.

This large toroidal magnetic confinement ITER machine is described from confinement process to advanced physics of plasma-wall interactions, where pulses erupt from core plasma blistering the machine walls. Emissions from the walls reduce the core temperature which must remain ten times hotter than the 15 million degree core solar temperature to maintain ITER fusion power. The huge temperature gradient from core to wall that drives intense plasma turbulence is described in detail.

Also explained are the methods designed to limit the growth of small magnetic islands, the growth of edge localized plasma plumes and the solid state physics limits of the stainless steel walls of the confinement vessel from the burning plasma. Designs of the wall coatings and the special "exhaust pipe" for spent hot plasma are provided in two chapters. And the issues associated with high-energy neutrons — about 10 times higher than in fission reactions — and how they are managed in ITER, are detailed.

Contents:

• Machine Architecture and Objectives
• Magnetohydrodynamic Description of the Equilibrium
• Alfv'en Cavity Modes, Fast Ions, Alpha Particles and Diagnostic Neutral Beams
• Turbulent Transport from the Temperature Gradients
• Operational Regimes and Their Properties
• Transport Barriers and ELM Control
• Steady-State Operation
• Plasma Diagnostics
• Plasma Facing Components and Plasma-Wall Interaction Physics
• The Broader Approach and Tritium Breeding Blankets

Readership: For nuclear fusion and ITER specialists.

Chapter 1

Machine Architecture and Objectives

1.1Beginning of the ITER Project

ITER began in 1985 as collaboration between four countries: Russia, the European Union (through the EURATOM organization), the USA, and Japan. Conceptual and engineering design phases were carried out under the auspices of the International Atomic Energy Agency (IAEA). By 2000 a design acceptable to all parties was completed. Subsequently, these four parties were joined by the People’s Republic of China and the Republic of South Korea. India became the seventh ITER partner in December 2005. The first Director General was Professor Kaname Ikeda and the second Director General was Professor Osamu Motojima. From 2015, the ITER Director is General Bernard Bigot from Commissariat Energie Atomic in France [http://www.cea.fr].

The first organized design activities (1980-1990) for a tokamak fusion reactor design were known as the INTOR project. Those first design activities also looked toward the second phase of a demonstration machine to follow the ITER proof-of-principle for fusion power. With the ITER project, the design efforts took a new direction. The earlier INTOR design activity is described by W. M. Stacey in The Quest for a Fusion Energy Reactor, Oxford, ISBN:978-0-19-973384-2.

In June 2005, the partners officially announced that ITER would be built in the European Union in Southern France. The negotiations that led to this decision ended in a compromise between the EU and Japan, in which Japan was promised directorship of 20% of the research staff at the French location of ITER, as well as having the Director of the administrative body of ITER. In addition, another research facility for the project was built in Japan, and the European Union contributes about 50% of the costs of this new Japanese-based institution located in Amori at the north end of Honshu. In November 2006 an international consortium signed a formal agreement to build the toroidal fusion machine. In September 2007, the People’s Republic of China became the seventh country to send their ITER Agreement to the IAEA. By October 2007, the ITER Agreement was established and the ITER Organization legally came into existence as a Treaty between the participating countries.

The initial design in the 1990s from the four partners is shown in Fig. 1.1 as presented by Aymar, et al. (1998). This design, called ITER-FDR, was for a major radius of 8 meters with the aim of achieving ignition and running at 1.5GW of fusion power. By 2000 the perspective had changed, owing to new results described in Chapter 6.

Fig. 1.1Cross-section of the original 1996 ITER-FDR architecture with R/a = 8.1m/2.80m and B/I_p = 5.7T/20MA designed for fusion power of 1.5GW. Here FDR refers to the Final Design Report from the Physics Expert Groups published in 1999. The international development path of this design is described in Stacey (2010). Subsequently, the design was charged to that shown in Fig. 1.2 with plasma current reduced to 15MA in view of lower-transport regimes discovered in the 2000s. Note the X-point of the separatrix line just above the dome of the divertor in the bottom of the inner chamber [Aymar, et al. (1998)]. The subsequent progress in improved plasma confinement described in Chapters 4-7 lead to the ITER-FEAT design [Campbell (2001)] with major radius R = 6.1m as shown in Fig. 1.2.

Construction of the ITER complex began in 2007 at a new site next to the Cadarache nuclear laboratory supported by the Commissaries Energie Atomique (CEA). The assembly of the tokamak building and machine was started in 2013-2015. The first components finished were the neutral beam injectors built in Japan. The construction of the ITER device and the supporting facilities is now (2015) well under way with the latest news available at https://www.iter.org.

1.2Architecture of ITER

The architecture of the ITER fusion reactor is shown in Fig. 1.2. The objectives of the engineering architecture are to design and build within a ten-year period, at reasonable cost, a tokamak capable of producing 500MW of output fusion power from 50MW of input power into a mixture of tritium and deuterium plasma. Achieving this power amplification factor Q_DT = 10 is considered more assured than the goal of reaching a self-sustained nuclear fusion from ignition, although the machine allows for the possibility of ignition if the thermal energy confinement proves to be sufficient. For a steady-state power system, the power amplification factor is traditionally defined as the Q of the system. ITER construction and research activities are constantly updated on the comprehensive website https://www.iter.org.

Fig. 1.2ITER (International Toroidal Experimental Reactor) architecture with R/a = 6.1m/2m with B₀ = 5.3T for confining plasma currents up to I_p = 15MA as given in Table 1.1. The table shows the change from the current European tokamak JET with new ITER fusion experiment.

The reference parameters for ITER, in comparison with the currently-operating Joint European Torus (JET) parameters, are shown in Table 1.1 from Shimada, et al. (2007). The JET deuterium-tritium fusion experiments in the 1997 time frame produced a power amplification factor of Q ~ 2/3 for pulse of about one second for 24MW injected and a power amplification of Q ~ 1/3 from a longer four-second pulse driven by 12MW of auxiliary heating power. Both the short and long pulses produced record amounts of fusion power, or equivalently a record number of neutrons from the DT nuclear fusion reactions. Thus, ITER may still be the first fusion confinement machine to show fusion power amplification factors greater than unity [Q > 1] for significant periods of time, meaning for time intervals of order tens of seconds. Reaching such net power amplification for a period time of order 50 to 100 energy confinement times is the key objective of ITER. Following this success a larger machine designed on what is learned from ITER will be used to develop the electric power producing fusion reactor called DEMO.

Table 1.1JET and ITER comparison of main parameters from [Shimada, et al. (2007)] in Physics Basis of ITER. In the left column are the best JET parameters as of 2015 and in the right column are the designed values for ITER.

	Maximum values achieved on JET separately	ITER Design values
R and a	3m 1.25m	6.2m 2m
Elongation	1.8	1.7
Plasma volume	100m^3	840m^3
Magnetic field on axis	4T	5.3 T
Plasma current in D-shaped plasma	7MA	16MA
Plateau current	1MA for 60 s	15MA for 500 s
Modes of operation	L, H and ELMy	ELMy
Plasma contact	carbon/beryllium	beryllium
	limiters-pumped divertor	pumped divertor
Neutral injection to the plasma	22MW	33-50MW
Coupled ICRH	22MW	20-40MW
ECRH	0	20-40MW
Current drive	3MA (LH) (5MA)	15MA
Solenoid	external	275V.s
Central density	2 × 10^20 m^−3	10^20 m^−3
Electron temperature	20 keV	21 keV
Ion temperature	40 keV	18 keV
Q value in DT plasma	0.6 (0.9 net)	10
Fusion power	16MW	500MW
Fusion energy	22 MJ in 4s	120 GJ in 200 s

The ITER machine is designed to produce a long DT fusion power pulse of up to 300 s with a Q = 10. The initial ITER design was given to the international fusion community [Aymar, et al. (1998)]. The machine architecture and the tokamak building and pit presented by Aymar and his design team are shown in Figs. 1.1. Note the components labeled divertor, divertor port, limiter, vacuum vessel, and blanket. These components are common to all fusion reactors. Following the improved confinement results obtained with internal transport barriers in the Japanese tokamak JT-60U reported by Fujita, et al. (1998, 1999) the ITER parameters were fixed with radii R/a = 6.2m 2.0m and B_T = 5.3T as given in the second column of Table 1.1. The second major ITER group publication is Campbell, et al. (2001).

Campbell (2001) describes the final design of the ITER machine with major radius R/a = 6.2m/2.0m and minor radius a = 2m and the mean toroidal magnetic field B = 5.3 T. The expectation is to achieve a plasma current I_p = 15MA and a fusion power amplification of Q = 10 from fusion power of 500MW.

The historical background of the design of ITER begins from the definitive fusion power experiments on the Tokamak Fusion Test Reactor that delivered the first deuterium-tritium experiments in 1993 and ran more than 840 DT discharges. The TFTR experiments were concluded in 1997. The first tritium experimental results in this relatively simple circular cross-section tokamak were reported by Bell, et al. (1989) at the International Atomic Energy Agency Meeting in Nice, France. A comprehensive review of the achievements of the TFTR experiments is given by Hawryluk (1998). The DT experiments were successful, producing a fusion power Q_fus = 0.5 ± 0.13 for pulses lasting a few energy confinement times. The TFTR plasmas reached record ion temperatures of 32 keV and thus high values of the fusion triple product of n_e(0)τ_ET_i(0) = 4 × 10²⁰ m³s keV by using high-power neutral beam injection [Hawryluk (1998)]. These first DT experiments were performed in a hot ion mode called supershots as described in Chapter 6. The TFTR discharges, or “shots” as referred to in the laboratory, produced 10-12MW of fusion power from injection of 40MW of NBI power over period of half a second. The experiments were important for showing that in the fusion plasma there are alpha particle-driven MHD modes, or instabilities, that are excited by the high-energy (3.5MeV) products of the fusion reactions [Nazikian, et al. (1997)]. Plasma instabilities excited by the 3.5MeV alpha particles released in the fusion reactions place significant constraints on the ITER system as described in Chapter 2. The instabilities in the core plasma show up as structures called “fishbones” and “sawteeth”, as explained in Chapter 3. The results of the TFTR experiments were largely incorporated in the design of the Joint European Tokamak or JET. JET began operation in 1991 [Jet Team (1992)] and achieved record fusion power gain Q_DT described in Keilhacker, et al. (1999).

A non-technical review of the history of the International Thermonuclear Experimental Reactor (ITER) is given by McCray (2010). McCray emphasizes three aspects of the project’s history, focusing largely on the European research community’s perspective. First, McCray explores how European scientists and science managers constructed a trans-national research community around fusion energy projects as part of Europe’s larger technological integration and development. McCray (2010) expands on Gabrielle Hecht’s concept of ‘technopolitics’ to the larger international dimension and explores how the political environments of the Cold War and the post-9/11 era helped shape ITER’s history, sometimes in ways not entirely within researchers’ control. The essay considers ITER as a technological project that gradually became globalized. At various stages in the project national borders became less important, while social, economic, legal and technological linkages created a shared ...

Table of contents

• Cover Page
• Title Page
• Copyright Page
• Prologue
• Contents
• 1. Machine Architecture and Objectives
• 2. Magnetohydrodynamic Description of the Equilibrium and Heating of the Thermal Plasma
• 3. Alfvén Cavity Modes, Fast Ions, Alpha Particles and Diagnostic Neutral Beams
• 4. Turbulent Transport from the Temperature Gradients
• 5. Operational Regimes and their Properties
• 6. Transport Barriers and ELM Control
• 7. Steady-State Operation
• 8. Plasma Diagnostics
• 9. Plasma Facing Components and Plasma-Wall Interaction Physics
• 10. The Broader Approach and Tritium Breeding Blankets
• Glossary Index
• General Index