CRCP-FUS-001

180 PP = 9.31 MM SPINE

INTERNATIONAL ATOMIC ENERGY AGENCY

VIENNA

World Survey of Fusion Devices 2022

WORLD SURVEY OF FUSION DEVICES

2022

AFGHANISTAN

ALBANIA

ALGERIA

ANGOLA

ANTIGUA AND BARBUDA

ARGENTINA

ARMENIA

AUSTRALIA

AUSTRIA

AZERBAIJAN

BAHAMAS

BAHRAIN

BANGLADESH

BARBADOS

BELARUS

BELGIUM

BELIZE

BENIN

BOLIVIA, PLURINATIONAL

STATE OF

BOSNIA AND HERZEGOVINA

BOTSWANA

BRAZIL

BRUNEI DARUSSALAM

BULGARIA

BURKINA FASO

BURUNDI

CAMBODIA

CAMEROON

CANADA

CENTRAL AFRICAN

REPUBLIC

CHAD

CHILE

CHINA

COLOMBIA

COMOROS

CONGO

COSTA RICA

CÔTE D’IVOIRE

CROATIA

CUBA

CYPRUS

CZECH REPUBLIC

DEMOCRATIC REPUBLIC

OF THE CONGO

DENMARK

DJIBOUTI

DOMINICA

DOMINICAN REPUBLIC

ECUADOR

EGYPT

EL SALVADOR

ERITREA

ESTONIA

ESWATINI

ETHIOPIA

FIJI

FINLAND

FRANCE

GABON

GEORGIA

GERMANY

GHANA

GREECE

GRENADA

GUATEMALA

GUYANA

HAITI

HOLY SEE

HONDURAS

HUNGARY

ICELAND

INDIA

INDONESIA

IRAN, ISLAMIC REPUBLIC OF

IRAQ

IRELAND

ISRAEL

ITALY

JAMAICA

JAPAN

JORDAN

KAZAKHSTAN

KENYA

KOREA, REPUBLIC OF

KUWAIT

KYRGYZSTAN

LAO PEOPLE’S DEMOCRATIC

REPUBLIC

LATVIA

LEBANON

LESOTHO

LIBERIA

LIBYA

LIECHTENSTEIN

LITHUANIA

LUXEMBOURG

MADAGASCAR

MALAWI

MALAYSIA

MALI

MALTA

MARSHALL ISLANDS

MAURITANIA

MAURITIUS

MEXICO

MONACO

MONGOLIA

MONTENEGRO

MOROCCO

MOZAMBIQUE

MYANMAR

NAMIBIA

NEPAL

NETHERLANDS

NEW ZEALAND

NICARAGUA

NIGER

NIGERIA

NORTH MACEDONIA

NORWAY

OMAN

PAKISTAN

PALAU

PANAMA

PAPUA NEW GUINEA

PARAGUAY

PERU

PHILIPPINES

POLAND

PORTUGAL

QATAR

REPUBLIC OF MOLDOVA

ROMANIA

RUSSIAN FEDERATION

RWANDA

SAINT KITTS AND NEVIS

SAINT LUCIA

SAINT VINCENT AND

THE GRENADINES

SAMOA

SAN MARINO

SAUDI ARABIA

SENEGAL

SERBIA

SEYCHELLES

SIERRA LEONE

SINGAPORE

SLOVAKIA

SLOVENIA

SOUTH AFRICA

SPAIN

SRI LANKA

SUDAN

SWEDEN

SWITZERLAND

SYRIAN ARAB REPUBLIC

TAJIKISTAN

THAILAND

TOGO

TONGA

TRINIDAD AND TOBAGO

TUNISIA

TÜRKİYE

TURKMENISTAN

UGANDA

UKRAINE

UNITED ARAB EMIRATES

UNITED KINGDOM OF

GREAT BRITAIN AND

NORTHERN IRELAND

UNITED REPUBLIC

OF TANZANIA

UNITED STATES OF AMERICA

URUGUAY

UZBEKISTAN

VANUATU

VENEZUELA, BOLIVARIAN

REPUBLIC OF

VIET NAM

YEMEN

ZAMBIA

ZIMBABWE

The following States are Members of the International Atomic Energy Agency:

The Agency’s Statute was approved on 23 October 1956 by the Conference on the Statute of the

IAEA held at United Nations Headquarters, New York; it entered into force on 29 July 1957.

The Headquarters of the Agency are situated in Vienna. Its principal objective is “to accelerate and enlarge

the contribution of atomic energy to peace, health and prosperity throughout the world’’.

WORLD SURVEY OF FUSION DEVICES

2022

INTERNATIONAL ATOMIC ENERGY AGENCY

VIENNA, 2022

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For further information on this publication, please contact:

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Printed by the IAEA in Austria

December 2022

IAEA Library Cataloguing in Publication Data

Names: International Atomic Energy Agency.

Title: World survey of fusion devices 2022 / International Atomic Energy Agency.

Description: Vienna : International Atomic Energy Agency, 2022. | Includes bibliographical

references.

Identifiers: IAEAL 22-01543 | ISBN 978–92–0–143422–7 (paperback : alk. paper) |

ISBN 978–92–0–143122–6 (pdf) |

Subjects: LCSH: | Nuclear fusion. | Nuclear fusion—Research. | Plasma (Ionized gases).

Classification: UDC 533.9 | CRCP/FUS/001

FOREWORD

The IAEA fosters discussion on nuclear fusion research and development advancing extensive

international dialogue to overcome highly technical challenges and make fusion energy a

reality. The IAEA is at the forefront of worldwide efforts to make fusion energy production a

reality, facilitating international coordination and sharing best practices in global projects.

In 2020 the IAEA released its first on-line database of fusion devices, the Fusion Device

Information System (FusDIS). FusDIS contains information on public and private fusion

devices with experimental and demonstration designs currently in operation, under construction

or being planned, as well as the technical data of these devices.

This publication compiles and further elaborates on the information available on FusDIS,

providing an up-to-date worldwide survey of experimental fusion devices, testing facilities and

plant designs.

The IAEA gratefully acknowledges the International Fusion Research Council for contributing

to this publication. The IAEA officer responsible for this publication was M. Barbarino of the

Division of Physical and Chemical Sciences.

EDITORIAL NOTE

This publication has been prepared from the original material as submitted by the contributors and has not been edited by the editorial

staff of the IAEA. The views expressed remain the responsibility of the contributors and do not necessarily represent the views of the

IAEA or its Member States.

Neither the IAEA nor its Member States assume any responsibility for consequences which may arise from the use of this publication.

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The use of particular designations of countries or territories does not imply any judgement by the publisher, the IAEA, as to the legal

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The mention of names of specific companies or products (whether or not indicated as registered) does not imply any intention to

infringe proprietary rights, nor should it be construed as an endorsement or recommendation on the part of the IAEA.

The IAEA has no responsibility for the persistence or accuracy of URLs for external or third party Internet web sites referred to in this

publication and does not guarantee that any content on such web sites is, or will remain, accurate or appropriate.

TABLE OF CONTENTS

1. INTRODUCTION ................................................................................................... 1

1.1. BACKGROUND ....................................................................................... 1

1.2. OBJECTIVE .............................................................................................. 7

1.3. SCOPE ....................................................................................................... 7

1.4. STRUCTURE ............................................................................................ 8

2. EXPERIMENTAL TOKAMAKS ......................................................................... 19

2.1. NOVA-FURG (FEDERAL UNIVERSITY OF ESPÍRITO SANTO,

BRAZIL) ................................................................................................. 19

2.1.1. Introduction .................................................................................. 19

2.1.2. Purpose ......................................................................................... 19

2.1.3. Main features ............................................................................... 19

2.2. ETE (NATIONAL INSTITUTE FOR SPACE RESEARCH, BRAZIL) 20

2.2.1. Introduction .................................................................................. 20

2.2.2. Purpose ......................................................................................... 20

2.2.3. Main features ............................................................................... 20

2.3. TCABR (UNIVERSITY OF SÃO PAULO, BRAZIL) .......................... 21

2.3.1. Introduction .................................................................................. 21

2.3.2. Purpose ......................................................................................... 21

2.3.3. Main features ............................................................................... 21

2.4. STOR-M (UNIVERSITY OF SASKATCHEWAN, CANADA) ........... 22

2.4.1. Introduction .................................................................................. 22

2.4.2. Purpose ......................................................................................... 22

2.4.3. Main features ............................................................................... 22

2.5. EAST (CHINESE ACADEMY OF SCIENCES, CHINA) ..................... 23

2.5.1. Introduction .................................................................................. 23

2.5.2. Purpose ......................................................................................... 23

2.5.3. Main features ............................................................................... 23

2.6. EXL-50 (ENN, CHINA) .......................................................................... 24

2.6.1. Introduction .................................................................................. 24

2.6.2. Purpose ......................................................................................... 24

2.6.3. Main features ............................................................................... 24

2.7. HL-2A (SOUTHWESTERN INSTITUTE OF PHYSICS, CHINA) ...... 25

2.7.1. Introduction .................................................................................. 25

2.7.2. Purpose ......................................................................................... 25

2.7.3. Main features ............................................................................... 25

2.8. HL-2M (SOUTHWESTERN INSTITUTE OF PHYSICS, CHINA) ..... 26

2.8.1. Introduction .................................................................................. 26

2.8.2. Purpose ......................................................................................... 26

2.8.3. Main features ............................................................................... 26

2.9. J-TEXT (HUAZHONG UNIVERSITY OF SCIENCE AND

TECHNOLOGY, CHINA) ...................................................................... 27

2.9.1. Introduction .................................................................................. 27

2.9.2. Purpose ......................................................................................... 27

2.9.3. Main features ............................................................................... 27

2.10. SUNIST-1 (TSINGHUA UNIVERSITY, CHINA) ................................ 28

2.10.1. Introduction .................................................................................. 28

2.10.2. Purpose ......................................................................................... 28

2.10.3. Main features ............................................................................... 28

2.11. MEDUSA-CR (INSTITUTO TECNOLOGICO DE COSTA RICA,

COSTA RICA) ........................................................................................ 29

2.11.1. Introduction .................................................................................. 29

2.11.2. Purpose ......................................................................................... 29

2.11.3. Main features ............................................................................... 29

2.12. GOLEM (CZECH TECHNICAL UNIVERSITY, CZECH

REPUBLIC) ............................................................................................. 30

2.12.1. Introduction .................................................................................. 30

2.12.2. Purpose ......................................................................................... 30

2.12.3. Main features ............................................................................... 30

2.13. COMPASS-U (INSTITUTE OF PLASMA PHYSICS, CZECH

REPUBLIC) ............................................................................................. 31

2.13.1. Introduction .................................................................................. 31

2.13.2. Purpose ......................................................................................... 31

2.13.3. Main features ............................................................................... 31

2.14. NORTH (TECHNICAL UNIVERSITY OF DENMARK,

DENMARK) ............................................................................................ 32

2.14.1. Introduction .................................................................................. 32

2.14.2. Purpose ......................................................................................... 32

2.14.3. Main features ............................................................................... 32

2.15. EGYPTOR (EGYPTIAN ATOMIC ENERGY AUTHORITY,

EGYPT) ................................................................................................... 33

2.15.1. Introduction .................................................................................. 33

2.15.2. Purpose ......................................................................................... 33

2.15.3. Main features ............................................................................... 33

2.16. ITER (ITER ORGANIZATION, FRANCE) ........................................... 34

2.16.1. Introduction .................................................................................. 34

2.16.2. Purpose ......................................................................................... 34

2.16.3. Main features ............................................................................... 34

2.17. WEST (CEA, FRANCE) ......................................................................... 35

2.17.1. Introduction .................................................................................. 35

2.17.2. Purpose ......................................................................................... 35

2.17.3. Main features ............................................................................... 35

2.18. ASDEX UPGRADE (MAX PLANCK INSTITUTE FOR PLASMA

PHYSICS, GERMANY) ......................................................................... 36

2.18.1. Introduction .................................................................................. 36

2.18.2. Purpose ......................................................................................... 36

2.18.3. Main features ............................................................................... 36

2.19. ADITYA-U (INSTITUTE FOR PLASMA RESEARCH, INDIA) ........ 37

2.19.1. Introduction .................................................................................. 37

2.19.2. Purpose ......................................................................................... 37

2.19.3. Main features ............................................................................... 37

2.20. SST-1 (INSTITUTE FOR PLASMA RESEARCH, INDIA) .................. 38

2.20.1. Introduction .................................................................................. 38

2.20.2. Purpose ......................................................................................... 38

2.20.3. Main features ............................................................................... 38

2.21. SSST (INSTITUTE FOR PLASMA RESEARCH, INDIA) ................... 39

2.21.1. Introduction .................................................................................. 39

2.21.2. Purpose ......................................................................................... 39

2.21.3. Main features ............................................................................... 39

2.22. ALVAND (IRAN ATOMIC ENERGY ORGANIZATION, ISLAMIC

REPUBLIC OF IRAN) ............................................................................ 40

2.22.1. Introduction .................................................................................. 40

2.22.2. Purpose ......................................................................................... 40

2.22.3. Main features ............................................................................... 40

2.23. DAMAVAND (IRAN ATOMIC ENERGY ORGANIZATION,

ISLAMIC REPUBLIC OF IRAN) .......................................................... 41

2.23.1. Introduction .................................................................................. 41

2.23.2. Purpose ......................................................................................... 41

2.23.3. Main features ............................................................................... 41

2.24. IR-T1 (ISLAMIC AZAD UNIVERSITY, ISLAMIC REPUBLIC OF

IRAN) ...................................................................................................... 42

2.24.1. Introduction .................................................................................. 42

2.24.2. Purpose ......................................................................................... 42

2.24.3. Main features ............................................................................... 42

2.25. DTT (ENEA, ITALY) ............................................................................. 43

2.25.1. Introduction .................................................................................. 43

2.25.2. Purpose ......................................................................................... 43

2.25.3. Main features ............................................................................... 43

2.26. FTU (ENEA, ITALY) ............................................................................. 44

2.26.1. Introduction .................................................................................. 44

2.26.2. Purpose ......................................................................................... 44

2.26.3. Main features ............................................................................... 44

2.27. LATE (KYOTO UNIVERSITY, JAPAN) .............................................. 45

2.27.1. Introduction .................................................................................. 45

2.27.2. Purpose ......................................................................................... 45

2.27.3. Main features ............................................................................... 45

2.28. PLATO (KYUSHU UNIVERSITY, JAPAN) ......................................... 46

2.28.1. Introduction .................................................................................. 46

2.28.2. Purpose ......................................................................................... 46

2.28.3. Main features ............................................................................... 46

2.29. QUEST (KYUSHU UNIVERSITY, JAPAN) ......................................... 47

2.29.1. Introduction .................................................................................. 47

2.29.2. Purpose ......................................................................................... 47

2.29.3. Main features ............................................................................... 47

2.30. HYBTOK-II (NAGOYA UNIVERSITY, JAPAN) ................................ 48

2.30.1. Introduction .................................................................................. 48

2.30.2. Purpose ......................................................................................... 48

2.30.3. Main features ............................................................................... 48

2.31. TOKASTAR-2 (NAGOYA UNIVERSITY, JAPAN) ............................ 49

2.31.1. Introduction .................................................................................. 49

2.31.2. Purpose ......................................................................................... 49

2.31.3. Main features ............................................................................... 49

2.32. JT-60SA (NATIONAL INSTITUTES FOR QUANTUM AND

RADIOLOGICAL SCIENCE AND TECHNOLOGY, JAPAN) ............ 50

2.32.1. Introduction .................................................................................. 50

2.32.2. Purpose ......................................................................................... 50

2.32.3. Main features ............................................................................... 50

2.33. TST-2 (THE UNIVERSITY OF TOKYO, JAPAN) ............................... 51

2.33.1. Introduction .................................................................................. 51

2.33.2. Purpose ......................................................................................... 51

2.33.3. Main features ............................................................................... 51

2.34. UTST (THE UNIVERSITY OF TOKYO, JAPAN) ............................... 52

2.34.1. Introduction .................................................................................. 52

2.34.2. Purpose ......................................................................................... 52

2.34.3. Main features ............................................................................... 52

2.35. PHIX (TOKYO INSTITUTE OF TECHNOLOGY, JAPAN) ................ 53

2.35.1. Introduction .................................................................................. 53

2.35.2. Purpose ......................................................................................... 53

2.35.3. Main features ............................................................................... 53

2.36. HIST (UNIVERSITY OF HYOGO, JAPAN) ......................................... 54

2.36.1. Introduction .................................................................................. 54

2.36.2. Purpose ......................................................................................... 54

2.36.3. Main features ............................................................................... 54

2.37. KTM (INSTITUTE OF ATOMIC ENERGY OF NATIONAL

NUCLEAR CENTER OF THE REPUBLIC OF KAZAKHSTAN,

KAZAKHSTAN) ..................................................................................... 55

2.37.1. Introduction .................................................................................. 55

2.37.2. Purpose ......................................................................................... 55

2.37.3. Main features ............................................................................... 55

2.38. LIBTOR (TAJOURA NUCLEAR RESEARCH CENTRE, LIBYA) .... 56

2.38.1. Introduction .................................................................................. 56

2.38.2. Purpose ......................................................................................... 56

2.38.3. Main features ............................................................................... 56

2.39. GLAST-III (PAKISTAN ATOMIC ENERGY COMMISSION,

PAKISTAN) ............................................................................................ 57

2.39.1. Introduction .................................................................................. 57

2.39.2. Purpose ......................................................................................... 57

2.39.3. Main features ............................................................................... 57

2.40. MT-1 (PAKISTAN ATOMIC ENERGY COMMISSION,

PAKISTAN) ............................................................................................ 58

2.40.1. Introduction .................................................................................. 58

2.40.2. Purpose ......................................................................................... 58

2.40.3. Main features ............................................................................... 58

2.41. MT-2 (PAKISTAN ATOMIC ENERGY COMMISSION,

PAKISTAN) ............................................................................................ 59

2.41.1. Introduction .................................................................................. 59

2.41.2. Purpose ......................................................................................... 59

2.41.3. Main features ............................................................................... 59

2.42. PST (PAKISTAN ATOMIC ENERGY COMMISSION, PAKISTAN). 60

2.42.1. Introduction .................................................................................. 60

2.42.2. Purpose ......................................................................................... 60

2.42.3. Main features ............................................................................... 60

2.43. ISTTOK (INSTITUTO SUPERIOR TÉCNICO, PORTUGAL) ............ 61

2.43.1. Introduction .................................................................................. 61

2.43.2. Purpose ......................................................................................... 61

2.43.3. Main features ............................................................................... 61

2.44. KSTAR (KOREA INSTITUTE OF FUSION ENERGY, REPUBLIC

OF KOREA) ............................................................................................ 62

2.44.1. Introduction .................................................................................. 62

2.44.2. Purpose ......................................................................................... 62

2.44.3. Main features ............................................................................... 62

2.45. VEST (SEOUL NATIONAL UNIVERSITY, REPUBLIC OF

KOREA) .................................................................................................. 63

2.45.1. Introduction .................................................................................. 63

2.45.2. Purpose ......................................................................................... 63

2.45.3. Main features ............................................................................... 63

2.46. FT-2 (IOFFE INSTITUTE, RUSSIAN FEDERATION) ........................ 64

2.46.1. Introduction .................................................................................. 64

2.46.2. Purpose ......................................................................................... 64

2.46.3. Main features ............................................................................... 64

2.47. GLOBUS-M2 (IOFFE INSTITUTE, RUSSIAN FEDERATION) ......... 65

2.47.1. Introduction .................................................................................. 65

2.47.2. Purpose ......................................................................................... 65

2.47.3. Main features ............................................................................... 65

2.48. TUMAN-3M (IOFFE INSTITUTE, RUSSIAN FEDERATION) .......... 66

2.48.1. Introduction .................................................................................. 66

2.48.2. Purpose ......................................................................................... 66

2.48.3. Main features ............................................................................... 66

2.49. T-15MD (NATIONAL RESEARCH CENTRE KURCHATOV

INSTITUTE, RUSSIAN FEDERATION) .............................................. 67

2.49.1. Introduction .................................................................................. 67

2.49.2. Purpose ......................................................................................... 67

2.49.3. Main features ............................................................................... 67

2.50. GUTTA (SAINT PETERSBURG STATE UNIVERSITY, RUSSIAN

FEDERATION) ....................................................................................... 68

2.50.1. Introduction .................................................................................. 68

2.50.2. Purpose ......................................................................................... 68

2.50.3. Main features ............................................................................... 68

2.51. T-11M (TROITSK INSTITUTE FOR INNOVATION AND FUSION

RESEARCH, RUSSIAN FEDERATION) .............................................. 69

2.51.1. Introduction .................................................................................. 69

2.51.2. Purpose ......................................................................................... 69

2.51.3. Main features ............................................................................... 69

2.52. SMART (UNIVERSITY OF SEVILLE, SPAIN) ................................... 70

2.52.1. Introduction .................................................................................. 70

2.52.2. Purpose ......................................................................................... 70

2.52.3. Main features ............................................................................... 70

2.53. TCV (SWISS PLASMA CENTER, SWITZERLAND) .......................... 71

2.53.1. Introduction .................................................................................. 71

2.53.2. Purpose ......................................................................................... 71

2.53.3. Main features ............................................................................... 71

2.54. TT-1 (THAILAND INSTITUTE OF NUCLEAR TECHNOLOGY,

THAILAND) ........................................................................................... 72

2.54.1. Introduction .................................................................................. 72

2.54.2. Purpose ......................................................................................... 72

2.54.3. Main features ............................................................................... 72

2.55. JET (EUROFUSION, UNITED KINGDOM) ......................................... 73

2.55.1. Introduction .................................................................................. 73

2.55.2. Purpose ......................................................................................... 73

2.55.3. Main features ............................................................................... 73

2.56. ST40 (TOKAMAK ENERGY, UNITED KINGDOM) .......................... 74

2.56.1. Introduction .................................................................................. 74

2.56.2. Purpose ......................................................................................... 74

2.56.3. Main features ............................................................................... 74

2.57. MAST-U (UKAEA, UNITED KINGDOM) ........................................... 75

2.57.1. Introduction .................................................................................. 75

2.57.2. Purpose ......................................................................................... 75

2.57.3. Main features ............................................................................... 75

2.58. HBT-EP (COLUMBIA UNIVERSITY, UNITED STATES OF

AMERICA) ............................................................................................. 76

2.58.1. Introduction .................................................................................. 76

2.58.2. Purpose ......................................................................................... 76

2.58.3. Main features ............................................................................... 76

2.59. SPARC (COMMONWEALTH FUSION SYSTEMS, UNITED

STATES OF AMERICA) ........................................................................ 77

2.59.1. Introduction .................................................................................. 77

2.59.2. Purpose ......................................................................................... 77

2.59.3. Main features ............................................................................... 77

2.60. DIII-D (GENERAL ATOMICS, UNITED STATES OF AMERICA) ... 78

2.60.1. Introduction .................................................................................. 78

2.60.2. Purpose ......................................................................................... 78

2.60.3. Main features ............................................................................... 78

2.61. LTX-Β (PRINCETON PLASMA PHYSICS LABORATORY,

UNITED STATES OF AMERICA) ........................................................ 79

2.61.1. Introduction .................................................................................. 79

2.61.2. Purpose ......................................................................................... 79

2.61.3. Main features ............................................................................... 79

2.62. NSTX-U (PRINCETON PLASMA PHYSICS LABORATORY,

UNITED STATES OF AMERICA) ........................................................ 80

2.62.1. Introduction .................................................................................. 80

2.62.2. Purpose ......................................................................................... 80

2.62.3. Main features ............................................................................... 80

2.63. PEGASUS-III (UNIVERSITY OF WISCONSIN-MADISON,

UNITED STATES OF AMERICA) ........................................................ 81

2.63.1. Introduction .................................................................................. 81

2.63.2. Purpose ......................................................................................... 81

2.63.3. Main features ............................................................................... 81

3. EXPERIMENTAL STELLARATORS/HELIOTRONS ....................................... 83

3.1. CFQS (SOUTHWEST JIAOTONG UNIVERSITY, CHINA) ............... 83

3.1.1. Introduction .................................................................................. 83

3.1.2. Purpose ......................................................................................... 83

3.1.3. Main features ............................................................................... 83

3.2. SCR-1 (INSTITUTO TECNOLOGICO DE COSTA RICA, COSTA

RICA) ...................................................................................................... 84

3.2.1. Introduction .................................................................................. 84

3.2.2. Purpose ......................................................................................... 84

3.2.3. Main features ............................................................................... 84

3.3. RENAISSANCE FUSION (RENAISSANCE FUSION, FRANCE) ...... 85

3.3.1. Introduction .................................................................................. 85

3.3.2. Purpose ......................................................................................... 85

3.3.3. Main features ............................................................................... 85

3.4. WENDELSTEIN 7-X (MAX PLANK INSTITUTE FOR PLASMA

PHYSICS, GERMANY) ......................................................................... 86

3.4.1. Introduction .................................................................................. 86

3.4.2. Purpose ......................................................................................... 86

3.4.3. Main features ............................................................................... 86

3.5. TJ-K (UNIVERSITY OF STUTTGART, GERMANY) ......................... 87

3.5.1. Introduction .................................................................................. 87

3.5.2. Purpose ......................................................................................... 87

3.5.3. Main features ............................................................................... 87

3.6. HELIOTRON J (KYOTO UNIVERSITY, JAPAN) ............................... 88

3.6.1. Introduction .................................................................................. 88

3.6.2. Purpose ......................................................................................... 88

3.6.3. Main features ............................................................................... 88

3.7. LHD (NATIONAL INSTITUTE FOR FUSION SCIENCE, JAPAN) ... 89

3.7.1. Introduction .................................................................................. 89

3.7.2. Purpose ......................................................................................... 89

3.7.3. Main features ............................................................................... 89

3.8. TJ-II (CIEMAT, SPAIN) ......................................................................... 90

3.8.1. Introduction .................................................................................. 90

3.8.2. Purpose ......................................................................................... 90

3.8.3. Main features ............................................................................... 90

3.9. URAGAN-2M (INSTITUTE OF PLASMA PHYSICS NATIONAL

SCIENCE CENTER, UKRAINE) ........................................................... 91

3.9.1. Introduction .................................................................................. 91

3.9.2. Purpose ......................................................................................... 91

3.9.3. Main features ............................................................................... 91

3.10. URAGAN-3M (INSTITUTE OF PLASMA PHYSICS NATIONAL

SCIENCE CENTER, UKRAINE) ........................................................... 92

3.10.1. Introduction .................................................................................. 92

3.10.2. Purpose ......................................................................................... 92

3.10.3. Main features ............................................................................... 92

3.11. CTH (AUBURN UNIVERSITY, UNITED STATES OF AMERICA) .. 93

3.11.1. Introduction .................................................................................. 93

3.11.2. Purpose ......................................................................................... 93

3.11.3. Main features ............................................................................... 93

3.12. HIDRA (UNIVERSITY OF ILLINOIS, UNITED STATES OF

AMERICA) ............................................................................................. 94

3.12.1. Introduction .................................................................................. 94

3.12.2. Purpose ......................................................................................... 94

3.12.3. Main features ............................................................................... 94

3.13. HSX (UNIVERSITY OF WISCONSIN-MADISON, UNITED

STATES OF AMERICA) ........................................................................ 95

3.13.1. Introduction .................................................................................. 95

3.13.2. Purpose ......................................................................................... 95

3.13.3. Main features ............................................................................... 95

4. EXPERIMENTAL INERTIAL/LASER FUSION DEVICES .............................. 97

4.1. HB11 (HB11 ENERGY, AUSTRALIA) ................................................. 97

4.1.1. Introduction .................................................................................. 97

4.1.2. Purpose ......................................................................................... 97

4.1.3. Main features ............................................................................... 97

4.2. LMJ (CEA, FRANCE) ............................................................................ 98

4.2.1. Introduction .................................................................................. 98

4.2.2. Purpose ......................................................................................... 98

4.2.3. Main features ............................................................................... 98

4.3. MARVEL FUSION (MARVEL FUSION, GERMANY) ....................... 99

4.3.1. Introduction .................................................................................. 99

4.3.2. Purpose ......................................................................................... 99

4.3.3. Main features ............................................................................... 99

4.4. GEKKO XII (OSAKA UNIVERSITY, JAPAN) .................................. 100

4.4.1. Introduction ................................................................................ 100

4.4.2. Purpose ....................................................................................... 100

4.4.3. Main features ............................................................................. 100

4.5. LFEX (OSAKA UNIVERSITY, JAPAN) ............................................ 101

4.5.1. Introduction ................................................................................ 101

4.5.2. Purpose ....................................................................................... 101

4.5.3. Main features ............................................................................. 101

4.6. FIRST LIGHT (FIRST LIGHT FUSION LTD, UNITED

KINGDOM) ........................................................................................... 102

4.6.1. Introduction ................................................................................ 102

4.6.2. Purpose ....................................................................................... 102

4.6.3. Main features ............................................................................. 102

4.7. INNOVEN ENERGY LLC (INNOVEN ENERGY, UNITED

STATES OF AMERICA) ...................................................................... 103

4.7.1. Introduction ................................................................................ 103

4.7.2. Purpose ....................................................................................... 103

4.7.3. Main features ............................................................................. 103

4.8. NIF (LAWRENCE LIVERMORE NATIONAL LABORATORY,

UNITED STATES OF AMERICA) ...................................................... 104

4.8.1. Introduction ................................................................................ 104

4.8.2. Purpose ....................................................................................... 104

4.8.3. Main features ............................................................................. 104

4.9. OMEGA (UNIVERSITY OF ROCHESTER LABORATORY FOR

LASER ENERGETICS, UNITED STATES OF AMERICA) .............. 105

4.9.1. Introduction ................................................................................ 105

4.9.2. Purpose ....................................................................................... 105

4.9.3. Main features ............................................................................. 105

5. EXPERIMENTAL ALTERNATIVE DEVICE CONCEPTS ............................. 107

5.1. GENERAL FUSION (GENERAL FUSION INC, CANADA) ............ 107

5.1.1. Introduction ................................................................................ 107

5.1.2. Purpose ....................................................................................... 107

5.1.3. Main features ............................................................................. 107

5.2. KTX (UNIVERSITY OF SCIENCE AND TECHNOLOGY OF

CHINA, CHINA) ................................................................................... 108

5.2.1. Introduction ................................................................................ 108

5.2.2. Purpose ....................................................................................... 108

5.2.3. Main features ............................................................................. 108

5.3. TORIX (ÉCOLE POLYTECHNIQUE, FRANCE) .............................. 109

5.3.1. Introduction ................................................................................ 109

5.3.2. Purpose ....................................................................................... 109

5.3.3. Main features ............................................................................. 109

5.4. RFX (CONSORZIO RFX, ITALY) ...................................................... 110

5.4.1. Introduction ................................................................................ 110

5.4.2. Purpose ....................................................................................... 110

5.4.3. Main features ............................................................................. 110

5.5. RELAX (KYOTO INSTITUTE OF TECHNOLOGY, JAPAN) .......... 111

5.5.1. Introduction ................................................................................ 111

5.5.2. Purpose ....................................................................................... 111

5.5.3. Main features ............................................................................. 111

5.6. UH-CTI (KYUSHU UNIVERSITY, JAPAN) ...................................... 112

5.6.1. Introduction ................................................................................ 112

5.6.2. Purpose ....................................................................................... 112

5.6.3. Main features ............................................................................. 112

5.7. FAT-CM (NIHON UNIVERSITY, JAPAN) ........................................ 113

5.7.1. Introduction ................................................................................ 113

5.7.2. Purpose ....................................................................................... 113

5.7.3. Main features ............................................................................. 113

5.8. RT-1 (THE UNIVERSITY OF TOKYO, JAPAN) ............................... 114

5.8.1. Introduction ................................................................................ 114

5.8.2. Purpose ....................................................................................... 114

5.8.3. Main features ............................................................................. 114

5.9. UH-MCPG1 (UNIVERSITY OF HYOGO, JAPAN) ........................... 115

5.9.1. Introduction ................................................................................ 115

5.9.2. Purpose ....................................................................................... 115

5.9.3. Main features ............................................................................. 115

5.10. GAMMA 10/PDX (UNIVERSITY OF TSUKUBA, JAPAN) ............. 116

5.10.1. Introduction ................................................................................ 116

5.10.2. Purpose ....................................................................................... 116

5.10.3. Main features ............................................................................. 116

5.11. PILOT GAMMA PDX-SC (UNIVERSITY OF TSUKUBA, JAPAN) 117

5.11.1. Introduction ................................................................................ 117

5.11.2. Purpose ....................................................................................... 117

5.11.3. Main features ............................................................................. 117

5.12. CAT (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN

FEDERATION) ..................................................................................... 118

5.12.1. Introduction ................................................................................ 118

5.12.2. Purpose ....................................................................................... 118

5.12.3. Main features ............................................................................. 118

5.13. GDMT (BUDKER INSTITUTE OF NUCLEAR PHYSICS,

RUSSIAN FEDERATION) ................................................................... 119

5.13.1. Introduction ................................................................................ 119

5.13.2. Purpose ....................................................................................... 119

5.13.3. Main features ............................................................................. 119

5.14. GDMT CORE (BUDKER INSTITUTE OF NUCLEAR PHYSICS,

RUSSIAN FEDERATION) ................................................................... 120

5.14.1. Introduction ................................................................................ 120

5.14.2. Purpose ....................................................................................... 120

5.14.3. Main features ............................................................................. 120

5.15. GDT (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN

FEDERATION) ..................................................................................... 121

5.15.1. Introduction ................................................................................ 121

5.15.2. Purpose ....................................................................................... 121

5.15.3. Main features ............................................................................. 121

5.16. GOL-NB (BUDKER INSTITUTE OF NUCLEAR PHYSICS,

RUSSIAN FEDERATION) ................................................................... 122

5.16.1. Introduction ................................................................................ 122

5.16.2. Purpose ....................................................................................... 122

5.16.3. Main features ............................................................................. 122

5.17. SMOLA (BUDKER INSTITUTE OF NUCLEAR PHYSICS,

RUSSIAN FEDERATION) ................................................................... 123

5.17.1. Introduction ................................................................................ 123

5.17.2. Purpose ....................................................................................... 123

5.17.3. Main features ............................................................................. 123

5.18. EXTRAP T2R (KTH ROYAL INSTITUTE OF TECHNOLOGY,

SWEDEN) ............................................................................................. 124

5.18.1. Introduction ................................................................................ 124

5.18.2. Purpose ....................................................................................... 124

5.18.3. Main features ............................................................................. 124

5.19. TORPEX (SWISS PLASMA CENTER, SWITZERLAND) ................ 125

5.19.1. Introduction ................................................................................ 125

5.19.2. Purpose ....................................................................................... 125

5.19.3. Main features ............................................................................. 125

5.20. FUSION POWER CORE (COMPACT FUSION SYSTEMS,

UNITED STATES OF AMERICA) ...................................................... 126

5.20.1. Introduction ................................................................................ 126

5.20.2. Purpose ....................................................................................... 126

5.20.3. Main features ............................................................................. 126

5.21. IDCD (CTFUSION, UNITED STATES OF AMERICA) .................... 127

5.21.1. Introduction ................................................................................ 127

5.21.2. Purpose ....................................................................................... 127

5.21.3. Main features ............................................................................. 127

5.22. HELICITY DRIVE (HELICITYSPACE, UNITED STATES OF

AMERICA) ........................................................................................... 128

5.22.1. Introduction ................................................................................ 128

5.22.2. Purpose ....................................................................................... 128

5.22.3. Main features ............................................................................. 128

5.23. POLARIS (HELION ENERGY, UNITED STATES OF AMERICA) . 129

5.23.1. Introduction ................................................................................ 129

5.23.2. Purpose ....................................................................................... 129

5.23.3. Main features ............................................................................. 129

5.24. TRENTA (HELION ENERGY, UNITED STATES OF AMERICA) .. 130

5.24.1. Introduction ................................................................................ 130

5.24.2. Purpose ....................................................................................... 130

5.24.3. Main features ............................................................................. 130

5.25. HORNE HYBRID REACTOR (HORNE TECHNOLOGIES LLC,

UNITED STATES OF AMERICA) ...................................................... 131

5.25.1. Introduction ................................................................................ 131

5.25.2. Purpose ....................................................................................... 131

5.25.3. Main features ............................................................................. 131

5.26. PJMIF (HYPERJET FUSION CORPORATION, UNITED STATES

OF AMERICA) ..................................................................................... 132

5.26.1. Introduction ................................................................................ 132

5.26.2. Purpose ....................................................................................... 132

5.26.3. Main features ............................................................................. 132

5.27. FOCUS FUSION (LAWRENCEVILLE PLASMA PHYSICS, INC.

DBA LPPFUSION, UNITED STATES OF AMERICA) ..................... 133

5.27.1. Introduction ................................................................................ 133

5.27.2. Purpose ....................................................................................... 133

5.27.3. Main features ............................................................................. 133

5.28. CFR (LOCKHEED MARTIN, UNITED STATES OF AMERICA) .... 134

5.28.1. Introduction ................................................................................ 134

5.28.2. Purpose ....................................................................................... 134

5.28.3. Main features ............................................................................. 134

5.29. MIFTI (MAGNETO-INERTIAL FUSION TECHNOLOGIES, INC.,

UNITED STATES OF AMERICA) ...................................................... 135

5.29.1. Introduction ................................................................................ 135

5.29.2. Purpose ....................................................................................... 135

5.29.3. Main features ............................................................................. 135

5.30. PFRC (PRINCETON FUSION SYSTEMS, UNITED STATES OF

AMERICA) ........................................................................................... 136

5.30.1. Introduction ................................................................................ 136

5.30.2. Purpose ....................................................................................... 136

5.30.3. Main features ............................................................................. 136

5.31. Z MACHINE (SANDIA NATIONAL LABORATORIES, UNITED

STATES OF AMERICA) ...................................................................... 137

5.31.1. Introduction ................................................................................ 137

5.31.2. Purpose ....................................................................................... 137

5.31.3. Main features ............................................................................. 137

5.32. NORMAN (C-2W) (TAE TECHNOLOGIES, UNITED STATES

OF AMERICA) ..................................................................................... 138

5.32.1. Introduction ................................................................................ 138

5.32.2. Purpose ....................................................................................... 138

5.32.3. Main features ............................................................................. 138

5.33. COPERNICUS (TAE TECHNOLOGIES, UNITED STATES OF

AMERICA) ........................................................................................... 139

5.33.1. Introduction ................................................................................ 139

5.33.2. Purpose ....................................................................................... 139

5.33.3. Main features ............................................................................. 139

5.34. ZEBRA (UNIVERSITY OF NEVADA, UNITED STATES OF

AMERICA) ........................................................................................... 140

5.34.1. Introduction ................................................................................ 140

5.34.2. Purpose ....................................................................................... 140

5.34.3. Main features ............................................................................. 140

5.35. MST (UNIVERSITY OF WISCONSIN-MADISON, UNITED

STATES OF AMERICA) ...................................................................... 141

5.35.1. Introduction ................................................................................ 141

5.35.2. Purpose ....................................................................................... 141

5.35.3. Main features ............................................................................. 141

5.36. FUZE-Q (ZAP ENERGY INC., UNITED STATES OF AMERICA) .. 142

5.36.1. Introduction ................................................................................ 142

5.36.2. Purpose ....................................................................................... 142

5.36.3. Main features ............................................................................. 142

6. DEMO DEVICES ................................................................................................ 143

6.1. CFETR (CHINESE CONSORTIUM, CHINA) .................................... 143

6.1.1. Introduction ................................................................................ 143

6.1.2. Purpose ....................................................................................... 143

6.1.3. Main features ............................................................................. 143

6.2. EU-DEMO (EUROFUSION, EUROPEAN UNION) ........................... 144

6.2.1. Introduction ................................................................................ 144

6.2.2. Purpose ....................................................................................... 144

6.2.3. Main features ............................................................................. 144

6.3. JA-DEMO (JAPANESE CONSORTIUM, JAPAN) ............................. 145

6.3.1. Introduction ................................................................................ 145

6.3.2. Purpose ....................................................................................... 145

6.3.3. Main features ............................................................................. 145

6.4. K-DEMO (KOREA INSTITUTE OF FUSION ENERGY,

REPUBLIC OF KOREA) ...................................................................... 146

6.4.1. Introduction ................................................................................ 146

6.4.2. Purpose ....................................................................................... 146

6.4.3. Main features ............................................................................. 146

6.5. DEMO-RF (RUSSIAN CONSORTIUM, RUSSIAN FEDERATION) 147

6.5.1. Introduction ................................................................................ 147

6.5.2. Purpose ....................................................................................... 147

6.5.3. Main features ............................................................................. 147

6.6. FDP (GENERAL FUSION INC, UNITED KINGDOM) ..................... 148

6.6.1. Introduction ................................................................................ 148

6.6.2. Purpose ....................................................................................... 148

6.6.3. Main features ............................................................................. 148

6.7. ST-E1 (TOKAMAK ENERGY, UNITED KINGDOM) ...................... 149

6.7.1. Introduction ................................................................................ 149

6.7.2. Purpose ....................................................................................... 149

6.7.3. Main features ............................................................................. 149

6.8. STEP (UKAEA, UNITED KINGDOM) ............................................... 150

6.8.1. Introduction ................................................................................ 150

6.8.2. Purpose ....................................................................................... 150

6.8.3. Main features ............................................................................. 150

6.9. ARC (COMMONWEALTH FUSION SYSTEMS, UNITED

STATES OF AMERICA) ...................................................................... 151

6.9.1. Introduction ................................................................................ 151

6.9.2. Purpose ....................................................................................... 151

6.9.3. Main features ............................................................................. 151

6.10. GA-FPP (GENERAL ATOMICS, UNITED STATES OF

AMERICA) ........................................................................................... 152

6.10.1. Introduction ................................................................................ 152

6.10.2. Purpose ....................................................................................... 152

6.10.3. Main features ............................................................................. 152

6.11. DA VINCI (TAE TECHNOLOGIES, UNITED STATES OF

AMERICA) ........................................................................................... 153

6.11.1. Introduction ................................................................................ 153

6.11.2. Purpose ....................................................................................... 153

6.11.3. Main features ............................................................................. 153

REFERENCES .............................................................................................................. 155

CONTRIBUTORS TO DRAFTING AND REVIEW .................................................. 161

1. INTRODUCTION

1.1.BACKGROUND

Nuclear fusion is the process by which two light atomic nuclei combine to form a single heavier

one while releasing massive amounts of energy.

Fusion reactions take place in a state of matter called plasma – a hot, charged gas made of

positive ions and free-moving electrons with unique properties distinct from solids, liquids or

gases.

The sun, along with all other stars, is powered by this reaction. To fuse in our sun, nuclei need

to collide with each other at extremely high temperatures, around ten million degrees Celsius.

The high temperature provides them with enough energy to overcome their mutual electrical

repulsion. Once the nuclei come within a very close range of each other, the attractive nuclear

force between them will outweigh the electrical repulsion and allow them to fuse. For this to

happen, the nuclei need to be confined within a small space to increase the chances of collision.

In the sun, the extreme pressure produced by its immense gravity creates the conditions for

fusion.

Ever since the theory of nuclear fusion was understood in the 1930s, scientists – and

increasingly also engineers – have been on a quest to recreate and harness it. That is because if

nuclear fusion can be replicated on earth at an industrial scale, it could provide virtually

limitless clean, safe, and affordable energy to meet the world’s energy demand.

Fusion generates four times more energy per kilogram of fuel than fission used in nuclear power

plants, and nearly four million times more energy than burning oil or coal.

Most of the fusion reactor concepts under development will use a mixture of deuterium and

tritium (or D-T) — hydrogen atoms that contain extra neutrons (Fig. 1). In theory, with just a

few grams of these reactants, it is possible to produce a terajoule of energy, which is

approximately the energy one person in a developed country needs over sixty years.

FIG.1. A mixture of deuterium and tritium – two Hydrogen isotopes – will be used to fuel future fusion

power plants. Inside the reactor, deuterium and tritium nuclei collide and fuse, releasing Helium and

neutrons.

Fusion fuel is plentiful and easily accessible: deuterium can be extracted inexpensively from

seawater, and tritium can potentially be produced from the reaction of fusion generated neutrons

with naturally abundant lithium. These fuel supplies would last for millions of years. Future

fusion reactors are also intrinsically safe and are not expected to produce high activity or long-

lived nuclear waste. Furthermore, as the fusion process is difficult to start and maintain, there

is no risk of a runaway reaction and meltdown.

Importantly, nuclear fusion – just like fission – does not emit carbon dioxide or other

greenhouse gases into the atmosphere, so it could be a long-term source of low-carbon

electricity from the second half of this century onwards.

While the sun’s massive gravitational force naturally induces fusion, without that force a

temperature even higher than in the sun is needed for the reaction to take place. On Earth, we

need temperatures of around 150 million degrees Celsius to make deuterium and tritium fuse,

while regulating pressure and magnetic forces at the same time, for a stable confinement of the

plasma and to maintain the fusion reaction long enough to produce more energy than what was

required to start the reaction.

While conditions that are very close to those required in a fusion reactor are now routinely

achieved in experiments, improved confinement properties and stability of the plasma are still

needed to maintain the reaction and produce energy in a sustained manner. Scientists and

engineers from all over the world continue to develop and test new materials and design new

technologies to achieve net fusion energy.

Nuclear fusion and plasma physics research are carried out in more than 50 countries, and fusion

reactions have been successfully produced in many experiments, albeit without so far

generating more energy than what was required to start the process. Experts have come up with

different designs and magnet-based machines in which fusion takes place, like stellarators and

tokamaks (Fig. 2), but also approaches that rely on lasers, linear devices and advanced fuels.

How long it will take for fusion energy to be successfully rolled out will depend on mobilizing

resources through global partnerships and collaboration, and on how fast the industry will be

able to develop, validate and qualify emerging fusion technologies. Another important issue is

to develop in parallel the necessary nuclear infrastructure, such as the requirements, standards

and good practices, relevant to the realization of this future energy source.

FIG.2. How a tokamak works: The electric field induced by a transformer drives a current (green

horizontal arrow) through the plasma column. This generates a poloidal magnetic field that bends the

plasma current into a circle (yellow vertical arrow). Bending the column into a circle prevents

leakage and doing this inside a doughnut-shaped vessel creates a vacuum. The other magnetic field

going around the length of the doughnut is referred to as toroidal (blue horizontal arrow). The

combination of these two fields creates a three-dimensional curve, like a helix (shown in black), in

which the plasma is highly confined. Twisting the magnets can also produce the helical shape without

the need for a transformer – this kind of configuration is called a stellarator.

Following 10 years of component design, site preparation, and manufacturing across the world,

the assembly of ITER in France, the world’s largest international fusion facility, commenced

in 2020. ITER (see p. 34) is an international project that aims to demonstrate the scientific and

technological feasibility of fusion energy production and prove technology and concepts for

future electricity-producing demonstration fusion power plants, called DEMOs (see pp. 143–

153). ITER will start conducting its first experiments in the second half of the 2020s and full

power experiments should commence in the second half of the 2030s.

DEMO timelines vary in different countries, but the consensus among experts is that an

electricity producing fusion power plant could be built and operating by 2050. In parallel,

numerous privately funded commercial enterprises are also making strides in developing

concepts for fusion power plants, drawing on the know-how generated over years of publicly

funded research and development, and proposing fusion power even sooner. Some of these

private companies are pursuing concepts based on fusion reactions other than the D-T reaction.

Traditionally, D-T reaction has been used to achieve fusion because they reach the highest

reaction rate at a lower temperature than other fuels. However, tritium is radioactive and does

not occur naturally in any significant quantities. Therefore, it has to be ‘bred’ in a nuclear

reaction between the fusion-generated neutrons and lithium surrounding the reactor wall. The

energy of these neutrons also presents significant challenges regarding the materials in the

reactor vacuum vessel, since, when the neutrons collide with the reactor walls, its structures

and components become radioactive. This necessitates considerations in radiation safety and

waste disposal.

To bypass the challenges caused by the use of tritium, there are now experiments using

alternative or advanced fusion fuels, like D-

He or p-

B. The list of the most favourable fusion

reactions is given in Table 1. Boron-11 is non-radioactive and comprises around 80 percent of

all boron found in nature, so it is readily available. However, the challenge with p-

B fusion

is that it would require the plasma to be a hundred times hotter than plasma containing

deuterium and tritium.

TABLE 1. LIST OF THE MOST FAVOURABLE FUSION REACTIONS

Reactants

Products

D-T

He (3.5 MeV) + n (14.1 MeV)

D-D

T (1.01 MeV) + p (3.02 MeV) (50%)

He (0.82 MeV) + n (2.45 MeV) (50%)

D-

He (3.6 MeV) + p (14.7 MeV)

T-T

He + 2 n + 11.3 MeV

He-

He + 2 p

He-T

He + p + n + 12.1 MeV (51%)

He (4.8 MeV) + D (9.5 MeV) (43%)

He (0.5 MeV) + n (1.9 MeV) + p (11.9 MeV) (6%)

D-

He + 22.4 MeV

He (1.7 MeV) +

He (2.3 MeV)

He-

He + p + 16.9 MeV

He + 8.7 MeV

Fusion research outputs have boomed over the last fifteen years since ITER was established,

by looking the number of first authorship papers presented at the IAEA Fusion Energy

Conference (FEC) between 2006 and 2021 (see Fig. 3). The United States of America (USA)

retains the top position with 152 first authorship papers for 2021. Japan and China are in second

and third place, respectively (see Fig. 4). The Princeton Plasma Physics Laboratory (USA) and

the Max Planck Institute for Plasma Physics (Germany) are the leading organizations in this

index, with a total of 42 first authorship papers in 2021 (see Fig. 5). The Institute for Plasma

Research (India) and Southwestern Institute of Physics (China) are not far behind, with 37 and

35 first authorship papers in 2021, respectively.

FIG.3. First authorship papers presented at the IAEA FEC between 2006 and 2021. Paper tracks are:

EX - Magnetic Fusion Experiments; TH - Magnetic Fusion Theory and Modelling; TECH - Fusion

Energy Technology; IFE - Inertial Fusion Energy; IAC - Innovative and Alternative Fusion Concepts;

OV – Overview.

FIG.4. Top ten countries per number of first authorship papers submitted at the 28th IAEA Fusion

Energy Conference (May 2021). Paper tracks are: EX - Magnetic Fusion Experiments; TH - Magnetic

Fusion Theory and Modelling; TECH - Fusion Energy Technology; IFE - Inertial Fusion Energy; IAC

- Innovative and Alternative Fusion Concepts; OV - Overview.

FIG.5. Top ten organizations per number of first authorship papers submitted at the 28th IAEA Fusion

Energy Conference (May 2021). Paper tracks are: EX - Magnetic Fusion Experiments; TH - Magnetic

Fusion Theory and Modelling; TECH - Fusion Energy Technology; IFE - Inertial Fusion Energy; IAC

- Innovative and Alternative Fusion Concepts; OV - Overview.

A similar increasing pattern emerges when looking at private sector investments over the past

two years. Over 30 private fusion companies can be found in Australia, Canada, China, France,

Germany, Israel, Italy, Japan, United Kingdom, and USA. As of July 2022, private sector

companies have declared that they have attracted around US$5 billion in total (Fig. 6).

FIG.6. Private sector companies have disclosed around US$5 billion in fusion funding (more than $3

billion since June 2021). Readapted and updated from: The chase for fusion energy, Nature (2021);

The global fusion industry in 2022, Fusion Industry Association (2022).

Reflecting the global interest rising around fusion energy R&D, this publication – based on the

IAEA’s Fusion Device Information System (FusDIS)

– provides information on all fusion

devices public or private with experimental and demonstration designs, which are currently in

operation, under construction or being planned, as well as technical data of these devices. An

overview is given in Fig. 7.

FIG.7. Over 130 experimental, public and private, fusion devices are operating, under construction or

being planned, while a number of organizations are considering designs for demonstration fusion

power plants.

1.2.OBJECTIVE

The objective of this publication is to provide a survey of public and private fusion devices

worldwide with experimental and demonstration designs, which are currently in operation,

under construction or being planned.

1.3. SCOPE

This publication focuses on the following device configuration categories:

• Tokamaks – both conventional and spherical type;

• Stellarators and heliotrons;

• Laser and inertial fusion; and

Available at https://nucleus.iaea.org/sites/fusionportal/Pages/FusDIS.aspx

• Alternative concepts – this category includes the following types: dense plasma focus; field

reversed configuration; inertial electrostatic fusion; levitated dipole; magnetic mirror

machine; magnetized target fusion; pinch; reverse field pinch; simple magnetized torus;

space propulsor; and spheromak.

For each device, the following information is provided:

• Country;

• Organization;

• Device name;

• Device type;

• Device status (operating; under construction; planned);

• Design (experimental; DEMO); and

• Ownership (public; private; public-private).

R&D in fusion science is also carried out at various experimental and testing facilities working

in the areas of plasma physics, material science, nuclear engineering, manufacturing and

robotics. These facilities have not been included in this publication.

1.4. STRUCTURE

This publication is divided into six parts:

 Section 1 (this section) gives a general background and describes the objective, scope and

structure of this publication;

 Section 2 features public and private experimental tokamaks, which are in operation,

under construction or being planned – see Table 2.

 Section 3 features public and private experimental stellarators and heliotrons, which are in

operation or being planned – see Table 3.

 Section 4 features public and private experimental inertial and laser based-fusion devices,

which are in operation or being planned – see Table 4.

 Section 5 features public and private experimental fusion devices based on alternative

confinement concepts, which are in operation, under construction or being planned – see

Table 5.

 Section 6 presents an overview of public and private fusion DEMO devices, which are

being planned – see Table 6.

TABLE 2. LIST OF EXPERIMENTAL TOKAMAKS

Country

Organization

Name

Type

Status

Design

Ownership

Brazil

Federal

University of

Espírito Santo

NOVA-

FURG

Conventional

Tokamak

Operating Experimental Public

National

Institute for

Space Research-

INPE

ETE

Spherical

Tokamak

Operating Experimental Public

University of

São Paulo

TCABR

Conventional

Tokamak

Operating Experimental Public

Canada

University of

Saskatchewan

STOR-M

Conventional

Tokamak

Operating Experimental Public

China

Chinese

Academy of

Sciences

EAST

Conventional

Tokamak

Operating Experimental Public

ENN EXL-50

Spherical

Tokamak

Operating Experimental Private

Southwestern

Institute of

Physics

HL-2A

Conventional

Tokamak

Operating Experimental Public

HL-2M

Conventional

Tokamak

Operating Experimental Public

Huazhong

University of

Science and

Technology

J-TEXT

Conventional

Tokamak

Operating Experimental Public

Tsinghua

University

SUNIST-1

Spherical

Tokamak

Operating Experimental Public

Costa

Rica

Instituto

Tecnologico de

Costa Rica

MEDUSA-

Spherical

Tokamak

Operating Experimental Public

Czech

Republic

Czech Technical

University

GOLEM

Conventional

Tokamak

Operating Experimental Public

Institute of

Plasma Physics

of the Czech

Academy of

Sciences

COMPASS-

Conventional

Tokamak

Under

construction

Experimental Public

Denmark

Technical

University of

Denmark

NORTH

Spherical

Tokamak

Operating Experimental Public

Egypt

Egyptian

Atomic Energy

Authority

EGYPTOR

Conventional

Tokamak

Operating Experimental Public

TABLE 2. LIST OF EXPERIMENTAL TOKAMAKS CONTINUED

Country

Organization

Name

Type

Status

Design

Ownership

France

CEA WEST

Conventional

Tokamak

Operating Experimental Public

ITER

Organization

ITER

Conventional

Tokamak

Under

construction

Experimental Public

Germany

Max Planck

Institute for

Plasma

Physics

ASDEX

UPGRADE

Conventional

Tokamak

Operating Experimental Public

India

Institute for

Plasma

Research

ADITYA-U

Conventional

Tokamak

Operating Experimental Public

SST-1

Conventional

Tokamak

Operating Experimental Public

SSST

Spherical

Tokamak

Under

construction

Experimental Public

Islamic

Republic

of Iran

Iran Atomic

Energy

Organization

ALVAND

Conventional

Tokamak

Operating Experimental Public

DAMAVAN

Conventional

Tokamak

Operating Experimental Public

Islamic Azad

University

IR-T1

Conventional

Tokamak

Operating Experimental Public

Italy ENEA

DTT

Conventional

Tokamak

Planned Experimental Public

FTU

Conventional

Tokamak

Operating Experimental Public

Japan

Kyoto

University

LATE

Spherical

Tokamak

Operating Experimental Public

Kyushu

University

PLATO

Conventional

Tokamak

Planned Experimental Public

QUEST

Spherical

Tokamak

Operating Experimental Public

Nagoya

University

HYBTOK-II

Conventional

Tokamak

Operating Experimental Public

TOKASTA

R-2

Conventional

Tokamak

Operating Experimental Public

National

Institutes for

Quantum and

Radiological

Science and

Technology

JT-60SA

Conventional

Tokamak

Operating Experimental Public

The

University of

Tokyo

TST-2

Spherical

Tokamak

Operating Experimental Public

UTST

Spherical

Tokamak

Operating Experimental Public

TABLE 2. LIST OF EXPERIMENTAL TOKAMAKS CONTINUED

Country

Organization

Name

Type

Status

Design

Ownership

Japan

Tokyo Institute

of Technology

PHiX

Conventional

Tokamak

Operating Experimental Public

University of

Hyogo

HIST

Spherical

Tokamak

Operating Experimental Public

Kazakhstan

Institute of

Atomic Energy

NNC RK

KTM

Spherical

Tokamak

Operating Experimental Public

Libya

Tajoura Nuclear

Research Centre

LIBTOR

Conventional

Tokamak

Operating Experimental Public

Pakistan

Pakistan Atomic

Energy

Commission

GLAST-

III

Spherical

Tokamak

Operating Experimental Public

MT-1

Spherical

Tokamak

Operating Experimental Public

MT-2

Spherical

Tokamak

Under

construction

Experimental Public

PST

Spherical

Tokamak

Planned Experimental Public

Portugal

Instituto

Superior

Técnico

ISTTOK

Conventional

Tokamak

Operating Experimental Public

Republic of

Korea

Korea Institute

of Fusion

Energy

KSTAR

Conventional

Tokamak

Operating Experimental Public

Seoul National

University

VEST

Spherical

Tokamak

Operating Experimental Public

Russian

Federation

Ioffe Institute

FT-2

Conventional

Tokamak

Operating Experimental Public

Globus-

Spherical

Tokamak

Operating Experimental Public

TUMAN

-3M

Conventional

Tokamak

Operating Experimental Public

National

Research Centre

Kurchatov

Institute

T-15MD

Conventional

Tokamak

Under

construction

Experimental Public

Saint Petersburg

State University

GUTTA

Spherical

Tokamak

Operating Experimental Public

Troitsk Institute

for Innovation

and Fusion

Research

T-11M

Conventional

Tokamak

Operating Experimental Public

TABLE 2. LIST OF EXPERIMENTAL TOKAMAKS CONTINUED

Country

Organization

Name

Type

Status

Design

Ownership

Spain

University of

Seville

SMART

Spherical

Tokamak

Planned Experimental Public

Switzerland

Swiss Plasma

Center

TCV

Conventional

Tokamak

Operating Experimental Public

Thailand

Institute of

Nuclear

Technology

TT-1

Conventional

Tokamak

Under

construction

Experimental Public

United

Kingdom

EUROfusion JET

Conventional

Tokamak

Operating Experimental Public

Tokamak

Energy

ST40

Spherical

Tokamak

Operating Experimental Private

UKAEA MAST-U

Spherical

Tokamak

Operating Experimental Public

United States

of America

Columbia

University

HBT-EP

Conventional

Tokamak

Operating Experimental Public

Commonwealth

Fusion Systems

SPARC

Conventional

Tokamak

Under

construction

Experimental Private

General

Atomics

DIII-D

Conventional

Tokamak

Operating Experimental Public

Princeton

Plasma Physics

Laboratory

LTX-β

Spherical

Tokamak

Operating Experimental Public

NSTX-U

Spherical

Tokamak

Operating Experimental Public

University of

Wisconsin-

Madison

PEGAS

US-III

Spherical

Tokamak

Operating Experimental Public

TABLE 3. LIST OF EXPERIMENTAL STELLARATORS AND HELIOTRONS

Country

Organization

Name

Type

Status

Design

Ownership

China

Southwest

Jiaotong

University

CFQS Stellarator Planned Experimental Public

Costa Rica

Instituto

Tecnologico De

Costa Rica

CSR-1 Stellarator Operating Experimental Public

France

Renaissance

Fusion

RENAISS

ANCE

FUSION

Stellarator Planned Experimental Private

Germany

Max Plank

Institute for

Plasma Physics

WENDEL

STEIN 7-

Stellarator Operating Experimental Public

University of

Stuttgart

TJ-K Stellarator Operating Experimental Public

Japan

National

Institute for

Fusion Science

LHD Heliotron Operating Experimental Public

Kyoto

University

HELIOTR

ON J

Heliotron Operating Experimental Public

Spain

CIEMAT

TJ-II

Stellarator

Operating

Experimental

Public

Ukraine

Institute of

Plasma Physics

National

Science Center

URAGAN

-2M

Stellarator Operating Experimental Public

URAGAN

-3M

Stellarator Operating Experimental Public

United

States of

America

Auburn

University

CTH Torsatron Operating Experimental Public

University of

Illinois

HIDRA

Stellarator/

Tokamak

Operating Experimental Public

University of

Wisconsin-

Madison

HSX Stellarator Operating Experimental Public

TABLE 4. LIST OF EXPERIMENTAL INERTIAL AND LASER FUSION DEVICES

Country

Organization

Name

Type

Status

Design

Ownership

Australia

HB11 Energy

HB11

Laser Fusion

Planned

Experimental

Private

France

CEA

LMJ

Laser Fusion

Operating

Experimental

Public

Germany Marvel Fusion

MARVEL

FUSION

Laser Fusion

Planned Experimental Private

Japan

Osaka

University

GEKKO

XII

Laser Fusion

Operating Experimental Public

LFEX

Laser Fusion

Operating

Experimental

Public

United

Kingdom

First Light

Fusion Ltd

FIRST

LIGHT

FUSION

Inertial

Fusion

Operating Experimental Private

United

States of

America

Innoven Energy

INNOVEN

ENERGY

LLC

Laser Fusion Planned Experimental Private

Lawrence

Livermore

National

Laboratory

NIF Laser Fusion Operating Experimental Public

University of

Rochester

Laboratory for

Laser

Energetics

OMEGA Laser Fusion Operating Experimental Public

TABLE 5. LIST OF EXPERIMENTAL ALTERNATIVE FUSION DEVICES

Country

Organization

Name

Type

Status

Design

Ownership

Canada

General Fusion

Inc

GENERAL

FUSION

Magnetized

Target

Fusion

Under

construction

Experimental Private

China

University Of

Science and

Technology of

China

KTX

Reversed

Field Pinch

Operating Experimental Public

France

École

Polytechnique

TORIX

Simple

Magnetized

Torus

Operating Experimental Public

Italy Consorzio RFX RFX

Reversed

Field Pinch

Operating Experimental Public

Japan

Kyoto Institute

of Technology

RELAX

Reversed

Field Pinch

Operating Experimental Public

Kyushu

University

UH-CTI Spheromak Operating Experimental Public

Nihon

University

FAT-CM

Field

Reversed

Config.

Operating Experimental Private

University of

Tokyo

RT-1

Levitated

Dipole

Operating Experimental Public

University of

Hyogo

UH-MCPG1 Spheromak Operating Experimental Public

University of

Tsukuba

GAMMA

10/PDX

Magnetic

Mirror

Machine

Operating Experimental Public

PILOT

GAMMA

PDX-SC

Magnetic

Mirror

Machine

Under

construction

Experimental Public

Russian

Federation

Budker Institute

of Nuclear

Physics

CAT

Magnetic

Mirror

Machine

Under

construction

Experimental Public

GDMT

Magnetic

Mirror

Machine

Planned Experimental Public

GDMT

CORE

Magnetic

Mirror

Machine

Planned Experimental Public

GDT

Magnetic

Mirror

Machine

Operating Experimental Public

GOL-NB

Magnetic

Mirror

Machine

Operating Experimental Public

SMOLA

Magnetic

Mirror

Machine

Operating Experimental Public

TABLE 5. LIST OF EXPERIMENTAL ALTERNATIVE FUSION DEVICES CONTINUED

Country

Organization

Name

Type

Status

Design

Ownership

Sweden

KTH Royal

Institute of

Technology

EXTRAP

T2R

Reversed

Field Pinch

Operating Experimental Public

United

States of

America

Compact

Fusion Systems

FUSION

POWER

CORE

Magnetized

Target

Fusion

Planned Experimental Private

CTFusion

IDCD

Spheromak

Operating

Experimental

Private

Helicity Space

HELICITY

DRIVE

Space

Propulsor

Planned Experimental Private

Helion Energy POLARIS

Field

Reversed

Config.

Planned Experimental Private

Helion Energy

Horne

Technologies

LLC

TRENTA

Field

Reversed

Config.

Operating Experimental Private

HORNE

HYBRID

REACTOR

Inertial

Electr.

Fusion

Under

construction

Experimental Private

Hyperjet Fusion

Corporation

PJMIF

Magnetized

Target

Fusion

Planned Experimental Private

Lawrenceville

Plasma Physics

FOCUS

FUSION

Dense

Plasma

Focus

Operating Experimental Private

Lockheed

Martin

CFR

Magnetic

Mirror

Machine

Operating Experimental Public

Magneto-

Inertial Fusion

Technologies

MIFTI Pinch Planned Experimental Private

Princeton

Fusion Systems

PFRC

Field

Reversed

Config.

Planned Experimental Private

Sandia National

Laboratories

MACHINE

Pinch Operating Experimental Public

TAE

Technologies

NORMAN

(C2-W)

Field

Reversed

Config.

Operating Experimental Private

COPERNIC

Field

Reversed

Config.

Under

construction

Experimental Private

University of

Nevada

ZEBRA Pinch Operating Experimental Public

University of

Wisconsin-

Madison

MST

Reversed

Field Pinch

Operating Experimental Private

Zap Energy Inc.

FUZE-Q

Pinch

Operating

Experimental

Private

TABLE 6. LIST OF DEMO DEVICES

Country

Organization

Name

Type

Status

Design

Ownership

China

Chinese

Consortium

CFETR

Conventional

Tokamak

Planned DEMO Public

European

Union

EUROfusion

EU-

DEMO

Conventional

Tokamak

Planned DEMO Public

Japan

Japanese

Consortium

J-DEMO

Conventional

Tokamak

Planned DEMO Public

Republic of

Korea

Korea Institute

of Fusion

Energy

K-DEMO

Conventional

Tokamak

Planned DEMO Public

Russian

Federation

Russian

Consortium

DEMO-RF

Conventional

Tokamak

Planned DEMO Public

United

Kingdom

General Fusion

Inc

FDP

Magnetized

Target Fusion

Planned DEMO

Public-

Private

Tokamak

Energy

ST-E1

Spherical

Tokamak

Planned DEMO Private

UKAEA STEP

Spherical

Tokamak

Planned DEMO Public

United

States of

America

Commonwealth

Fusion Systems

ARC

Conventional

Tokamak

Planned DEMO Private

General

Atomics

GA-FPP

Conventional

Tokamak

Planned DEMO Private

TAE

Technologies

DA VINCI

Field

Reversed

Config.

Planned DEMO Private

2. EXPERIMENTAL TOKAMAKS

2.1. NOVA-FURG (FEDERAL UNIVERSITY OF ESPÍRITO SANTO, BRAZIL)

2.1.1. Introduction

NOVA-FURG was formerly known as the NOVA-II tokamak and belonged to Kyoto

University, Japan. It has been operating in Brazil since 1996. It is a small device in comparison

with the other tokamaks around the world, but its size gives certain advantages, as for example

it is cheaper to operate, and thus it makes possible to repeat an experiment many times for

testing a theory.

2.1.2. Purpose

The purpose of NOVA-FURG tokamak is study plasma-wall interaction and optical diagnostic

development.

2.1.3. Main features

NOVA-FURG is a small iron-cored machine operating with conducting shell stabilization [1].

Technical information is listed in Table 7 below.

TABLE 7. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.3 m

Minor radius, a

0.06 m

Plasma current, I

0.01 MA

Toroidal field, B

1 T

Pulse length

0.015 s

Magnetic field configuration

Circular limiter

Ownership

Public

2.2. ETE (NATIONAL INSTITUTE FOR SPACE RESEARCH, BRAZIL)

2.2.1. Introduction

The ETE spherical tokamak became operational at the end of 2000. It was fully constructed at

Associated Plasma Laboratory of Brazil’s National Institute for Space Research, Brazil.

2.2.2. Purpose

The main objectives of the project are to study the basic physics of low aspect ratio geometry

plasmas with emphasis on the plasma edge as well as on plasma wall interaction. Development

of new diagnostics and training in tokamak operation are also important objectives of ETE.

2.2.3. Main features

The ETE spherical tokamak is a small-to-medium size low aspect ratio machine. ETE was

designed with a minimum set of coils for the poloidal and toroidal fields, minimizing stray

magnetic fields and preserving good plasma accessibility. All coils are water cooled and were

manufactured with standard pure cooper. The toroidal field system consists of 12 D-shaped

single coils connected in series by two feed rings that compensate the stray magnetic field [2].

Technical information is listed in Table 8 below.

TABLE 8. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.3 m

Minor radius, a

0.2 m

Plasma current, I

0.045 MA

Toroidal field, B

0.4 T

Pulse length

0.006–0.012 s

Magnetic field configuration

Graphite limiter

Ownership

Public

2.3. TCABR (UNIVERSITY OF SÃO PAULO, BRAZIL)

2.3.1. Introduction

Originally TCABR was designed by Swiss Federal Institute of Technology Lausanne,

Switzerland where it was operated from 1980 to 1992. Later it was rebuilt at the Plasma Physics

Laboratory of the University of São Paulo, Brazil with new systems of discharge control and of

Alfvén wave excitation and named TCABR. The first plasma was produced in 1999.

2.3.2. Purpose

The main purpose of TCABR is investigating plasma heating by Alfvén waves.

2.3.3. Main features

TCABR is a medium-sized, ohmically heated tokamak with a circular plasma cross-section that

works with the hydrogen plasma. TCABR has the main standard diagnostics such as microwave

interferometry, optical spectroscopy, soft and hard X ray detection, electron–cyclotron emission

detector, bolometer, H

-emission detector, Mirnov coils and a variety of magnetic and

electrostatic probes [3]. Technical information is listed in Table 9 below.

TABLE 9. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.61 m

Minor radius, a

0.18 m

Plasma current, I

0.1 MA

Toroidal field, B

1.07 T

Pulse length

0.1 s

Magnetic field configuration

Poloidal graphite limiter

Magnetic ergodic limiter

Ownership

Public

2.4. STOR-M (UNIVERSITY OF SASKATCHEWAN, CANADA)

2.4.1. Introduction

STOR-M is a research tokamak that was built at University of Saskatchewan, Canada. After

Tokamak de Varennes was closed in 1997, STOR-M became the only Canadian magnetic

fusion device.

2.4.2. Purpose

STOR-M aims to demonstrate the feasibility of quasi steady state tokamak reactors, study the

compact torus injection (a method of fuelling the core of tokamak fusion reactors), as well as

develop novel far infrared lasers-based diagnostics for ITER and research on plasma assisted

material synthesis and processing.

2.4.3. Main features

The STOR-M tokamak is a small iron core research tokamak. For heating it uses method based

on exploiting ohmic H-modes, turbulent heating. STOR-M is equipped with feedback control

system for horizontal and vertical plasma positions, a driver for fast rising ohmic current, a

circuit system for alternating current operation, compact torus injector, and diagnostics [4].

Technical information is listed in Table 10 below.

TABLE 10. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.46 m

Minor radius, a

0.125 m

Plasma current, I

30–60 MA

Toroidal field, B

0.5–1 T

Pulse length

0.002–0.005 s

Magnetic field configuration

Stainless steel limiter (circular)

Ownership

Public

2.5. EAST (CHINESE ACADEMY OF SCIENCES, CHINA)

2.5.1. Introduction

EAST is located at Institute of Plasma Physics Chinese Academy of Sciences (ASIPP), China.

It followed the first Chinese superconducting tokamak HT-7, built by ASIPP in partnership

with the Russian Federation in the early 1990s. The first plasma at EAST was obtained in 2006.

2.5.2. Purpose

EAST aims to contribute studies of plasma physics, provide a scientific basis for the design and

construction of experimental reactors, including ITER, as EAST has shape and equilibrium

similar to ITER.

2.5.3. Main features

EAST is a tokamak operating with superconducting magnets. The main distinguishing features

of EAST are its non-circular cross-section, fully superconducting magnets and fully actively

water-cooled plasma facing components, which will be beneficial to explore the advanced

steady-state plasma operation modes [5]. Technical information is listed in Table 11 below.

TABLE 11. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.7 m

Minor radius, a

0.4 m

Plasma current, Ip

1 MA

Toroidal field, B

3.5 T

Pulse length

1-1000 s

Magnetic field configuration

Double-null divertor

Pump limiter

Single null divertor

Ownership

Public

2.6. EXL-50 (ENN, CHINA)

2.6.1. Introduction

EXL-50 is the Chinese first medium-sized experimental spherical tokamak. It was built by the

ENN Group, a Chinese energy company. The first plasma discharge without solenoid was

obtained in 2019. The EXL-50 is also part of the ENN Compact Fusion Project, funded by the

private company ENN Group.

2.6.2. Purpose

The main purpose of EXL-50 is to simplify engineering requirements of a fusion device. One

of the key experimental objectives of the EXL-50 is to test the efficiency of the electron

cyclotron resonance heating and the current drive in the absence of the central solenoid magnet.

2.6.3. Main features

The EXL-50 device is a medium-sized experimental spherical tokamak with a cylindrical

vacuum vessel and with fully non-inductive current drive. EXL-50 has six poloidal field coils

which are located outside the vacuum vessel and the toroidal field coil conductors to obtain

higher toroidal field discharges [6]. Technical information is listed in Table 12 below.

TABLE 12. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.58 m

Minor radius, a

0.41 m

Plasma current, I

0.5 MA

Toroidal field, B

0.45 T

Pulse length

5 s

Magnetic field configuration

Copper limiters coated with 0.3 mm tungsten

Ownership

Public

2.7. HL-2A (SOUTHWESTERN INSTITUTE OF PHYSICS, CHINA)

2.7.1. Introduction

Built in 2002, HL-2A was the first tokamak with a divertor in China. The tokamak is based on

main components (magnet coils and plasma vessel) belonging to the former ASDEX tokamak.

HL-2A is located at Southwestern Institute of Physics, China.

2.7.2. Purpose

The key mission of the HL-2A tokamak programme is to address critical physics and

technology issues for ITER and next-step fusion devices. The research focuses on radio

frequency wave heating, current drive, plasma confinement, turbulent transport, MHD

instabilities, energetic particle physics, H-mode and ELM control.

2.7.3. Main features

HL-2A is a medium-sized tokamak with a lower single null divertor configuration. Recently a

2 MW lower hybrid current drive system with passive active multi-junction antenna has been

developed for HL-2A [7]. Technical information is listed in Table 13 below.

TABLE 13. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.65 m

Minor radius, a

0.4 m

Plasma current, I

0.48 MA

Toroidal field, B

2.8 T

Pulse length

5 s

Magnetic field configuration

Limiter

lower single null divertor

Ownership

Public

2.8. HL-2M (SOUTHWESTERN INSTITUTE OF PHYSICS, CHINA)

2.8.1. Introduction

HL-2M is located at Southwestern Institute of Physics, China. It is a new machine with the

highest plasma performance in China. The first plasma was achieved in 2020.

2.8.2. Purpose

The key areas of the HL-2M research are high performance and high beta scenarios compatible

with flexible advanced divertor configurations, tests and validation of high heat flux plasma-

facing components, as well as investigation of advanced plasma physics with high performance.

2.8.3. Main features

The HL-2M device is a tokamak with copper conductor magnets. It is equipped with a more

effective and flexible divertor and a new set of toroidal and poloidal field coils [8]. HL-2M

features advanced divertor configurations (snowflake, tripod). Technical information is listed

in Table 14 below.

TABLE 14. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.78 m

Minor radius, a

0.65 m

Plasma current, I

3 MA

Toroidal field, B

3 T

Pulse length

10 s

Magnetic field configuration

Divertor (snowflake, tripod, single null, double

null)

Ownership

Public

2.9. J-TEXT (HUAZHONG UNIVERSITY OF SCIENCE AND TECHNOLOGY, CHINA)

2.9.1. Introduction

The J-TEXT, formerly TEXT/TEXT-U at University of Texas at Austin, USA, is located at

Huazhong University of Science and Technology, China. First plasma was produced in 2007.

2.9.2. Purpose

The main purpose of J-TEXT is proving fundamental physics and control mechanisms of

fusion plasma confinement and stability in support of ITER’s successful operation and the

design of CFETR (see p. 143).

2.9.3. Main features

The J-TEXT is a conventional tokamak with an iron core. The original limiter configuration

(with three moveable poloidal rail limiter targets) was upgraded in 2016. J-TEXT configuration

makes it possible to study 3-D effects and disruption mitigation in a tokamak thanks to the

upgraded resonant magnetic perturbation system and the shattered pellet injection system [9].

Technical information is listed in Table 15 below.

TABLE 15. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.05 m

Minor radius, a

0.25–0.29 m

Plasma current, I

0.2 MA

Toroidal field, B

2 T

Pulse length

0.8 s

Magnetic field configuration

Titanium-carbide-coated graphite limiter

Divertor

Ownership

Public

2.10. SUNIST-1 (TSINGHUA UNIVERSITY, CHINA)

2.10.1. Introduction

SUNIST-1 is the first spherical tokamak built in China in a partnership between National Nature

Science Foundation, Tsinghua University and Institute of Physics, Chinese Academy of

Sciences. First plasma was produced in 2002 and the machine was subsequently upgraded in

2008.

2.10.2. Purpose

The research objectives of SUNIST-1 are to increase the understanding of toroidal plasma

physics with a low aspect ratio and to produce a maintainable target plasma with non-inductive

start-up.

2.10.3. Main features

SUNIST-1 is a small-scale spherical tokamak. The central stack is the most important

component of its design, consisting of a 1 mm thick 304 stainless steel column, surrounding the

ohmic solenoid and toroidal field inner limbs. In the recent years, SUNIST-1 has been upgraded

with newly developed plasma diagnostics and plasma actuators. In addition, a high current

density plasma gun was installed [10]. Technical information is listed in Table 16 below.

TABLE 16. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.3 m

Minor radius, a

0.23 m

Plasma current, I

0.05–1 MA

Toroidal field, B

0.15 T

Pulse length

0.001 s

Magnetic field configuration

Poloidal limiters

Ownership

Public

2.11. MEDUSA-CR (INSTITUTO TECNOLOGICO DE COSTA RICA, COSTA RICA)

2.11.1. Introduction

MEDUSA-CR is a low aspect ratio spherical tokamak built by University of Wisconsin-

Madison, USA and donated to Instituto Tecnológico de Costa Rica, Costa Rica.

2.11.2. Purpose

MEDUSA-CR is operated for educational and training activities in the field of plasma physics

and tokamak physics.

2.11.3. Main features

The vacuum vessel of MEDUSA-CR is made of stainless steel, with the external Alfvén wave

antennas and an ergodic limiter both placed externally to the vessel [11]. Technical information

is listed in Table 17 below.

TABLE 17. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.14 m

Minor radius, a

0.1 m

Plasma current, I

0.04 MA

Toroidal field, B

0.5 T

Pulse length

0.003 s

Magnetic field configuration

Rail limiter

Ownership

Public

2.12. GOLEM (CZECH TECHNICAL UNIVERSITY, CZECH REPUBLIC)

2.12.1. Introduction

Originally built in the 1960s (as TM-1 tokamak) at Kurchatov Institute, Russian Federation –

the GOLEM tokamak is located at Czech Technical University, Czech Republic. It is the oldest

tokamak in operation.

2.12.2. Purpose

The GOLEM tokamak is used for educational and training activities in the ﬁeld of tokamak

physics, technology, diagnostics and operation.

2.12.3. Main features

GOLEM is a small tokamak with a circular cross-section. Remote participation and control

using internet access are unique features of GOLEM, allowing for basic remote control in both

online and offline mode [12]. Technical information is listed in Table 18 below.

TABLE 18. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.4 m

Minor radius, a

0.085 m

Plasma current, I

0.008 MA

Toroidal field, B

0.8 T

Pulse length

0.025 s

Magnetic field configuration

Limiter

Ownership

Public

2.13. COMPASS-U (INSTITUTE OF PLASMA PHYSICS, CZECH REPUBLIC)

2.13.1. Introduction

COMPASS-U is a high magnetic field tokamak being constructed at Institute of Plasma

Physics, Czech Republic in cooperation between various European and international partners.

Operation is expected to start in 2023.

2.13.2. Purpose

COMPASS-U research will support ITER’s operation and help address key challenges in EU-

DEMO design activities (see DEMOs section). The device will allow for operation with hot

first wall and full recycling regime, and it is designed to study and test liquid metal technology

in the divertor.

2.13.3. Main features

The main features are a closed divertor with high plasma and neutral density, high opacity,

extreme power fluxes, high magnetic field, allowing access to advanced confinement modes

[13]. Technical information is listed in Table 19 below.

TABLE 19. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Under construction

Major radius, R

0.894

Minor radius, a

0.27

Plasma current, I

2 MA

Toroidal field, B

5 T

Pulse length

1–3 s (up to 11 s in lower single null plasmas)

Magnetic field configuration

Lower single null, negative triangularity with

limited plasma information (Phase 1-2)

Double null (Phase 2-3)

Snowflake, negative triangularity (Phase 3-4)

Ownership

Public

2.14. NORTH (TECHNICAL UNIVERSITY OF DENMARK, DENMARK)

2.14.1. Introduction

NORTH is a tokamak operating at Technical University of Denmark, Denmark, since 2019.

This tokamak is a joint project between the Technical University of Denmark and private sector

company Tokamak Energy, UK who constructed the device. The device is the first tokamak

experiment at the Technical University of Denmark [14].

2.14.2. Purpose

NORTH is used for plasma physics research.

2.14.3. Main features

The NORTH tokamak is a small-scale spherical tokamak with two magnetrons (3 kW each)

operating at 2.45 GHz being used for plasma heating. Technical information is listed in Table

20 below.

TABLE 20. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.25 m

Toroidal field, B

0.3 T

Ownership

Public

2.15. EGYPTOR (EGYPTIAN ATOMIC ENERGY AUTHORITY, EGYPT)

2.15.1. Introduction

EGYPTOR is a small tokamak originally built by Heinrich-Heine Universität Düsseldorf,

Germany and later re-installed in Egypt.

2.15.2. Purpose

EGYPTOR research programme aims at testing new diagnostic and cleaning discharge

techniques for developing better wall conditioning systems for large tokamaks.

2.15.3. Main features

EGYPTOR is a small-scale tokamak with stainless steel vacuum vessel, control and cleaning

discharge systems. Diagnostics can be accessed thanks to six large lateral ports and twelve

smaller windows located at top and bottom of the vessel [15]. Technical information is listed

in Table 21 below.

TABLE 21. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.3 m

Minor radius, a

0.1 m

Plasma current, I

0.05 MA

Toroidal field, B

1.2 T

Pulse length

0.045 s

Magnetic field configuration

Rail limiter

Ownership

Public

2.16. ITER (ITER ORGANIZATION, FRANCE)

2.16.1. Introduction

ITER is an experimental fusion reactor under construction in France. The project is a joint

international undertaking between China, the European Union with UK and Switzerland

through other agreements, India, Japan, South Korea, Russian Federation, and the USA. As of

May 2022, its construction is about 75% complete, with the successful testing of many key

components such as vacuum vessel sectors, toroidal field coils, poloidal field coils, cryostat

sections, and thermal shields. ITER will start conducting its first experiments in the second half

of 2020 and full-power experiments are planned to commence in 2036.

2.16.2. Purpose

The purpose of ITER is to demonstrate the scientific and technological feasibility of fusion

energy production. The principal scientific mission objectives of the ITER project are

demonstrating a scientific energy gain

sci

≥10 for deuterium-tritium plasma burn durations of

300–500 s (inductive ELMy H-mode); and the development of long-pulse, non-inductive

scenarios aiming at maintaining Q

sci

~5 for periods of up to 3000 s [16].

2.16.3. Main features

Weighing 23 000 tonnes and standing at nearly 30 metres tall, ITER will sit at the heart of a

180-hectare site, together with auxiliary housing and equipment. The ITER tokamak will have

a plasma volume of 830 m

. Its magnet system will be made of 18 toroidal field magnets, 6

poloidal field coils, a thirteen meters tall central solenoid, 18 superconducting correction coils,

31 superconducting magnet feeders and 29 non-superconducting in-vessel coils. The divertor

will be made up of 54 stainless-steel pieces known as cassettes, each weighing 10 tonnes. The

components facing the plasma will be armoured with tungsten, a material that has both low

tritium absorption and the highest melting temperature of any natural element. The ITER

external heating systems will rely on 33 MW neutral beam injection, 20 MW ion cyclotron

heating and 20 MW electron cyclotron heating. Technical information is listed in Table 22

below.

TABLE 22. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Under construction

Major radius, R

6.2 m

Minor radius, a

2 m

Plasma current, I

15 MA

Toroidal field, B

5.3 T

Pulse length

up to 3000 s

Magnetic field configuration

Divertor

Ownership

Public

The scientific energy gain corresponds to the fusion energy released (heat produced) divided by the energy

delivered to the fuel (heating given). This is different from the engineering energy gain (Q

eng

), which

corresponds to the ratio of grid power to recirculating electrical power. ITER is designed to achieve scientific

energy gain. DEMO-type devices are designed to achieve net engineering gain (Q

eng

>1).

2.17. WEST (CEA, FRANCE)

2.17.1. Introduction

Previously known as Tore Supra and renamed WEST following an upgrade in the configuration

(from limiter to divertor), this tokamak has been operating since 1988. WEST first plasma was

produced in 2016. WEST was the first tokamak with superconducting magnets and actively

cooled plasma facing components.

2.17.2. Purpose

Key missions of WEST are the qualification of the high heat flux plasma facing components in

integrating both technological and physics aspects in relevant heat and particle exhaust

conditions (particularly for the tungsten monoblocks foreseen in ITER divertor), as well as the

demonstration of integrated steady state operation, with a focus on power exhaust issues.

2.17.3. Main features

WEST is a superconducting tokamak equipped with two up-down symmetric divertors.

Different divertor configurations can be studied, i.e., lower single null, upper single null and

double null. The plasma facing components on WEST are made of tungsten and are actively

cooled. Thanks to the radiofrequency heating and current drive systems (9 MW of ion cyclotron

resonance heating and 6 MW of lower hybrid current drive), WEST can operate with long

pulses up to 1000 s with high particle fluence [17]. Technical information is listed in Table 23

below.

TABLE 23. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

2.5 m

Minor radius, a

0.5 m

Plasma current, I

1 MA

Toroidal field, B

3.7 T

Pulse length

Up to 1000 s

Magnetic field configuration

Divertor lower single null, upper single null and

double null

Ownership

Public

2.18. ASDEX UPGRADE (MAX PLANCK INSTITUTE FOR PLASMA PHYSICS,

GERMANY)

2.18.1. Introduction

ASDEX Upgrade (AUG) is a newly built tokamak which is based on the experience of the

ASDEX tokamak. AUG is operating since 1991.

2.18.2. Purpose

AUG programme aims to establish the scientific basis for the optimisation of the tokamak

approach to fusion energy and prepare for ITER and DEMO.

2.18.3. Main features

AUG is a medium sized tokamak with stainless steel vacuum vessel inner wall clad with tiles

made of or coated with tungsten metal. Sixteen large copper magnet coils wrapped around the

donut-shaped plasma vessel form the confining magnetic field. The device is equipped with

three different plasma heating methods: neutral particle injection (20 MW), high-frequency

heating (6 MW), and microwave heating (8 MW) [18]. Technical information is listed in Table

24 below.

TABLE 24. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.65 m

Minor radius, a

0.5 m

Plasma current, I

1.4 MA

Toroidal field, B

3.2 T

Pulse length

10 s

Magnetic field configuration

D-shaped divertor

Ownership

Public

2.19. ADITYA-U (INSTITUTE FOR PLASMA RESEARCH, INDIA)

2.19.1. Introduction

ADITYA-U is the upgraded version of ADITYA tokamak. First plasma in ADITYA-U was

obtained in 2016.

2.19.2. Purpose

The main purpose of ADITYA-U is to provide the physical and technological scientific base

for future fusion installations with special emphasis on performing experiments on disruption

and runaway electron mitigation.

2.19.3. Main features

ADITYA-U is a medium-sized tokamak. ADITYA-U is equipped with an open divertor

configuration, with divertor plates without any baffle and with three set of divertor coils. Two

sets of coils are placed at high field side and a set of coils is placed at the low field side to obtain

the shaped plasmas in lower single null, upper single null and double null divertor

configurations [19]. ADITYA-U is equipped with inductively driven macroparticle injector

system. Technical information is listed in Table 25 below.

TABLE 25. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.75 m

Minor radius, a

0.2–0.25 m

Plasma current, I

0.1–0.25 MA

Toroidal field, B

1.5 T

Pulse length

0.25–0.4 s

Magnetic field configuration

Circular and Single/Double null divertor

Ownership

Public

2.20. SST-1 (INSTITUTE FOR PLASMA RESEARCH, INDIA)

2.20.1. Introduction

Operating since 2013, SST-1 is located at Institute for Plasma Research, India.

2.20.2. Purpose

The focus of SST-1 research plan is to study plasma current drive through lower hybrid,

including the active feedback control and plasma-wall interactions in steady-state plasmas.

2.20.3. Main features

SST-1 is a medium-sized superconducting tokamak having superconducting toroidal field

magnets operating in two-phase helium in cryo-stable conditions. Recent modifications in the

external ohmic coils system have allowed repeatable and consistent ohmic plasmas with

electron cyclotron assisted pre-ionization [20]. Technical information is listed in Table 26

below.

TABLE 26. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.1 m

Minor radius, a

0.2 m

Plasma current, I

0.1 MA

Toroidal field, B

1.5 T (max 3 T)

Pulse length

0.65 s

Magnetic field configuration

Circular limiter

Ownership

Public

2.21. SSST (INSTITUTE FOR PLASMA RESEARCH, INDIA)

2.21.1. Introduction

Currently under construction, SSST will be the first spherical tokamak of the Institute for

Plasma Research, India.

2.21.2. Purpose

SSST research plan will focus on producing low aspect ratio plasmas and perform various basic

plasma experiments to study low aspect ratio plasmas, ohmic start-up with electron cyclotron

resonance heating and non-inductive start-up, among others. The device will also be used for

training in the field of low aspect ratio plasmas and associated technologies.

2.21.3. Main features

SSST will be a low cost spherical tokamak All coils will be made of copper and naturally

cooled. Toroidal field and poloidal field coils will be demountable and driven by capacitor bank

based power supplies. The general configuration of all power supplies will comprise of fast

and slow capacitor banks and electronic switches that discharge the energy stored in the banks

on the magnetic coil. The vacuum vessel will be made of SS304 and will be a continuous

structure without any toroidal breaks. Two magnetron (6 kW) based radiofrequency sources

will be used for pre-ionization and for non-inductive current drive during advanced operation

of the machine. The device will be equipped with several type of diagnostics such as Rogowski

coils, Langmuir probes, bolometers, microwave and infrared diagnostics and spectroscopy,

among others. Technical information is listed in Table 27 below.

TABLE 27. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Under construction

Major radius, R

0.28 m

Minor radius, a

0.16 m

Plasma current, I

0.028 MA

Toroidal field, B

0.15 T

Ownership

Public

2.22. ALVAND (IRAN ATOMIC ENERGY ORGANIZATION, ISLAMIC REPUBLIC OF

IRAN)

2.22.1. Introduction

The Alvand tokamak is one of the three tokamaks operating in the Islamic Republic of Iran.

2.22.2. Purpose

Alvand is used to measure plasma information and study its physical properties.

2.22.3. Main features

Alvand tokamak is a small size tokamak with circular plasma cross section [21]. Technical

information is listed in Table 28 below.

TABLE 28. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.456 m

Minor radius, a

0.126 m

Plasma current, I

0.035 MA

Toroidal field, B

0.8 T

Pulse length

0.009 s

Magnetic field configuration

Circular limiter

Ownership

Public

2.23. DAMAVAND (IRAN ATOMIC ENERGY ORGANIZATION, ISLAMIC REPUBLIC

OF IRAN)

2.23.1. Introduction

Damavand tokamak was built from Russian tokamak TVD, reducing plasma elongation from 4

to 2.

2.23.2. Purpose

Damavand research is devoted to studies of plasma discharges with magnetic configuration

similar to ITER tokamak.

2.23.3. Main features

Damavand is a small tokamak with an elongated plasma cross section and a poloidal divertor.

Its passive coils inside the vacuum chamber provide the plasma formation at the centre of the

torus, acting as a passive plasma current stabilizer [22]. Technical information is listed in Table

29 below.

TABLE 29. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.362 m

Minor radius, a

0.077 m

Plasma current, I

0.03 MA

Toroidal field, B

0.9 T

Pulse length

0.02 s

Ownership

Public

2.24. IR-T1 (ISLAMIC AZAD UNIVERSITY, ISLAMIC REPUBLIC OF IRAN)

2.24.1. Introduction

IR-T1 is based on a tokamak called HT-6B, which was originally built in China in 1984.

Currently IR-T1 is located at Plasma Physics Research Center of Islamic Azad University in

Tehran, Islamic Republic of Iran. First plasma was obtained in 1994.

2.24.2. Purpose

The main purpose of IR-T1 is studying plasma information under different experimental

conditions.

2.24.3. Main features

IR-T1 is a small tokamak with large aspect ratio and with a circular cross section. The tokamak

chamber is made of stainless steel and is surrounded by leaden walls. It consists of two poloidal

stainless-steel limiters with a radial thickness of 2.5 cm [23]. Technical information is listed in

Table 30 below.

TABLE 30. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.45 m

Minor radius, a

0.125 m

Plasma current, I

0.04 MA

Toroidal field, B

0.9 T

Pulse length

0.03 s

Magnetic field configuration

Ring limiter (tungsten)

Ownership

Public

2.25. DTT (ENEA, ITALY)

2.25.1. Introduction

The DTT is a tokamak being constructed at ENEA centre in Frascati, Italy. Operation is

expected to start by 2026.

2.25.2. Purpose

The main purpose of DTT is to study strategies for the management of plasma exhaust in a

reactor-grade tokamak plasma in support of ITER operation and DEMO design studies.

2.25.3. Main features

The DTT will be a superconducting tokamak. It will have a divertor configuration that can work

in single and double null scenarios, as well as a number of advanced divertor configurations.

The maximum performance of auxiliary heating power will be 45 MW [24]. Technical

information is listed in Table 31 below.

TABLE 31. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Under construction

Major radius, R

2.19 m

Minor radius, a

0.7 m

Plasma current, I

5.5 MA

Toroidal field, B

6 T

Pulse length

100 s

Magnetic field configuration

Divertor

Ownership

Public

2.26. FTU (ENEA, ITALY)

2.26.1. Introduction

Operating since 1990, FTU was the first tokamak that performed experiments with a liquid

lithium limiter.

2.26.2. Purpose

The main purpose of FTU tokamak is to study and optimize techniques of reduction of the

impurities in plasma using different elements, i.e., liquid metal first wall materials [39]. Most

of FTU experiments since 2018 are devoted to studies on liquid metal limiters, runaway

electrons and MHD stability.

2.26.3. Main features

FTU is a medium-sized tokamak with a high toroidal magnetic field, a circular poloidal cross

section and metallic first wall. The vacuum chamber is made of 2 mm thick stainless steel and

is lined internally with 2 cm thick toroidal molybdenum tile limiter. The machine also features

an external poloidal molybdenum limiter [25]. Technical information is listed in Table 32

below.

TABLE 32. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.935 m

Minor radius, a

0.3 m

Plasma current, I

0.04 MA

Toroidal field, B

8 T

Pulse length

1.5 s

Magnetic field configuration

Limiter

Ownership

Public

2.27. LATE (KYOTO UNIVERSITY, JAPAN)

2.27.1. Introduction

LATE is a fusion device located at Kyoto University, Japan.

2.27.2. Purpose

The purpose of the LATE device is to investigate and establish the physical bases on formation

of low aspect ratio torus plasmas by electron cyclotron heating and electron cyclotron current

drive solely.

2.27.3. Main features

LATE is a spherical tokamak without central solenoid. The non-inductive start-up is achieved

by using electron cyclotron heating and electron cyclotron current drive [26]. General

information is listed in Table 33 below.

TABLE 33. GENERAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Ownership

Public

2.28. PLATO (KYUSHU UNIVERSITY, JAPAN)

2.28.1. Introduction

PLATO is an experimental fusion device currently being developed and planned to be located

at Kyushu University, Japan.

2.28.2. Purpose

Research on PLATO will focus on the physics of turbulent plasmas and their understanding.

2.28.3. Main features

PLATO will be the first device able to measure the entire cross-sections of turbulent plasma

with a spatial resolution of microscale of Lamour radius. The device will allow to explore

plasma cross-scale couplings, turbulence localization and principles of structural formation and

function expression, by using tomography, heavy ion beam probe and microwaves diagnostics

[27]. Technical information is listed in Table 34 below.

TABLE 34. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Major radius, R

0.7 m

Minor radius, a

0.25 m

Plasma current, I

0.075 MA

Toroidal field, B

0.3 T

Pulse length

0.2 s

Ownership

Public

2.29. QUEST (KYUSHU UNIVERSITY, JAPAN)

2.29.1. Introduction

QUEST is a device located at Kyushu University, Japan. It was built in 2008, becoming the

largest spherical tokamak in Japan.

2.29.2. Purpose

The main purpose of QUEST research programme is to develop an integrated understanding of

particle balance in the plasma core, scrape-off layer, and plasma facing walls.

2.29.3. Main features

QUEST is a medium-sized spherical tokamak with two plasma heating sources with frequency

and power of 8.2 GHz, 50 kW and 28 GHz, 350 kW, respectively [28]. Technical information

is listed in Table 35 below.

TABLE 35. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.64 m

Minor radius, a

0.4 m

Plasma current, I

0.3 MA

Toroidal field, B

0.25 T

Ownership

Public

2.30. HYBTOK-II (NAGOYA UNIVERSITY, JAPAN)

2.30.1. Introduction

Built in 1997, HYBTOK-II is a tokamak built by and located at Nagoya University, Japan.

2.30.2. Purpose

HYBTOK-II was built to study particle transport characteristics in the edge plasma.

2.30.3. Main features

HYBTOK-II is a small conventional tokamak with a set of magnetic field coils installed to

perturbate the magnetic field confining the plasma from the outside [29]. Technical information

is listed in Table 36 below.

TABLE 36. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.4 m

Minor radius, a

0.128 m

Plasma current, I

0.015 MA

Toroidal field, B

0.5 T

Pulse length

0.01 s

Magnetic field configuration

Ergodic divertor

Ownership

Public

2.31. TOKASTAR-2 (NAGOYA UNIVERSITY, JAPAN)

2.31.1. Introduction

TOKASTAR-2 is device with a variable configuration, which can operate either as a tokamak

or a stellarator. TOKASTAR-2 is the successor of TOKASTAR and C-TOKASTAR devices

and operational since 2009.

2.31.2. Purpose

The main purpose of TOKASTAR-2 is to investigate the effects of outer helical field

application on tokamak plasmas.

2.31.3. Main features

On TOKASTAR-2, a set of magnetic coils can generate a variable configuration allowing to

operate the machine either as a tokamak or a stellarator, independently [30]. Technical

information is listed in Table 37 below.

TABLE 37. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.1 m

Minor radius, a

0.04 m

Plasma current, I

0.025 MA

Toroidal field, B

0.1 T

Ownership

Public

2.32. JT-60SA (NATIONAL INSTITUTES FOR QUANTUM AND RADIOLOGICAL

SCIENCE AND TECHNOLOGY, JAPAN)

2.32.1. Introduction

JT-60SA is the upgrade version of the JT-60U tokamak and is located at Naka Fusion Institute,

Japan. Its commissioning started in 2020 and it was interrupted due to insufficient voltage

insulation capability in one of the magnetic coils. Improvement for isolation capability is on-

going and integrated commissioning is expected to restart early in 2023. JT-60SA is a joint

project between Japan and the European Union.

2.32.2. Purpose

The main purpose of the JT-60SA project is to support fusion R&D by addressing key physics

and technology issues relevant for ITER operation and DEMO design studies.

2.32.3. Main features

JT-60SA is a fully superconducting tokamak capable of confining high temperature deuterium

plasmas. It can operate with both single and double null divertor configurations and with a wide

range of plasma shapes and aspect ratios [31]. The total heating power is of 41 MW. Technical

information is listed in Table 38 below.

TABLE 38. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating (under repair until 2023)

Major radius, R

2.96 m

Minor radius, a

1.18 m

Plasma current, I

5.5 MA

Toroidal field, B

2.25 T

Pulse length

100 s

Magnetic field configuration

Single and double null divertor

Ownership

Public

2.33. TST-2 (THE UNIVERSITY OF TOKYO, JAPAN)

2.33.1. Introduction

Operating since 1999, TST-2 is located at University of Tokyo, Japan. TST-2 is the upgraded

version of TST-M.

2.33.2. Purpose

The main purpose of TST-2 is to study helicity injection and turbulence-induced transport.

2.33.3. Main features

TST-2 is a medium-sized spherical tokamak. Recently manufactured, the vacuum vessel is

continuous in the toroidal direction and consists of a 1.4 m diameter, 6 mm thick stainless-steel

cylinder and top and bottom domes [32]. Technical information is listed in Table 39 below.

TABLE 39. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.36 m

Minor radius, a

0.23 m

Plasma current, I

0.2 MA

Toroidal field, B

0.3 T

Pulse length

0.05 s

Ownership

Public

2.34. UTST (THE UNIVERSITY OF TOKYO, JAPAN)

2.34.1. Introduction

UTST is a spherical tokamak located at University of Tokyo, Japan. Its construction started in

2004 and became operational in 2007.

2.34.2. Purpose

The main objectives of UTST are to study central solenoid-free start-up scheme and to obtain

high-beta spherical tokamak equilibria by using the plasma merging technique.

2.34.3. Main features

UTST is a medium-sized spherical tokamak without poloidal field coils [33]. Technical

information is listed in Table 40 below.

TABLE 40.TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.35 m

Minor radius, a

0.2 m

Plasma current, I

0.1 MA

Toroidal field, B

0.25 T

Pulse length

0.01 s

Ownership

Public

2.35. PHIX (TOKYO INSTITUTE OF TECHNOLOGY, JAPAN)

2.35.1. Introduction

PHiX is a tokamak located at Tokyo Institute of Technology, Japan, operational since 2014.

2.35.2. Purpose

The main purpose of PHiX is to study both longitudinal cross-section and vertical position

stability of fusion plasma [34].

2.35.3. Main features

PHiX is a small-sized tokamak. Technical information is listed in Table 41 below.

TABLE 41. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.33 m

Minor radius, a

0.09 m

Plasma current, I

0.005 MA

Toroidal field, B

0.3 T

Pulse length

0.02 s

Ownership

Public

2.36. HIST (UNIVERSITY OF HYOGO, JAPAN)

2.36.1. Introduction

HIST is a spherical tokamak located at University of Hyogo, Japan.

2.36.2. Purpose

The main purpose of HIST is to study the helicity injection physics on the spherical tokamak-

line.

2.36.3. Main features

HIST is a spherical tokamak with stainless-steel vacuum vessel, 9 mm thick, 1.5 m in diameter

and 3 m long [35]. Technical information is listed in Table 42 below.

TABLE 42. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.3 m

Minor radius, a

0.24 m

Plasma current, I

0.1 MA

Toroidal field, B

0.2 T

Ownership

Public

2.37. KTM (INSTITUTE OF ATOMIC ENERGY OF NATIONAL NUCLEAR CENTER

OF THE REPUBLIC OF KAZAKHSTAN, KAZAKHSTAN)

2.37.1. Introduction

The KTM tokamak is operating since 2017 at Institute of Atomic Energy of National Nuclear

Center of the Republic of Kazakhstan.

2.37.2. Purpose

The main purpose of the KTM tokamak is to study and test materials and design solutions for

protecting the fusion reactor first wall, to develop methods to reduce the heat loads on divertor

plates and divertor units as well as various methods of heat and energy removal, including ways

to quickly pump out the divertor volume, and develop methods for preventing out-of-order

failure of intra-chamber elements.

2.37.3. Main features

KTM is a spherical tokamak with aspect ratio equal to 2, plasma elongation κ

=1.7 and

radiofrequency heating power of 5–7 MW. Its divertor design, which consists of plasma-facing

plates mounted on a rotary table, is an asset for materials testing campaigns, allowing swiftly

replacements of the divertor plates [36]. Technical information is listed in Table 43 below.

TABLE 43. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.9 m

Minor radius, a

0.45 m

Plasma current, I

0.75 MA

Toroidal field, B

1 T

Pulse length

5 s

Magnetic field configuration

Single null divertor

Ownership

Public

2.38. LIBTOR (TAJOURA NUCLEAR RESEARCH CENTRE, LIBYA)

2.38.1. Introduction

The LIBTOR tokamak is located at Tajoura Nuclear Research Centre, Libya.

2.38.2. Purpose

The main purpose of LIBTOR research programme is to study plasma confinement and

transport models.

2.38.3. Main features

LIBTOR is a small-sized conventional tokamak. Technical information is listed in Table 44

below.

TABLE 44. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.53 m

Minor radius, a

0.115 m

Plasma current, I

0.12 MA

Toroidal field, B

4 T

Magnetic field configuration

Rail limiter

Ownership

Public

2.39. GLAST-III (PAKISTAN ATOMIC ENERGY COMMISSION, PAKISTAN)

2.39.1. Introduction

GLAST-III, located in Pakistan, Islamabad, is the upgrade from GLAST-I and GLAST-II

tokamaks. These devices were developed by Pakistan Atomic Energy Commission, National

Tokamak Fusion Programme of Pakistan.

2.39.2. Purpose

The main purpose of GLAST-III is to study tokamak plasma in a small dielectric vacuum vessel

with electron cyclotron resonance-assisted tokamak start-up.

2.39.3. Main features

GLAST-III is a small-sized spherical tokamak with a vacuum vessel made of Pyrex glass.

Technical information is listed in Table 45 below.

TABLE 45. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.2 m

Minor radius, a

0.1 m

Plasma current, I

0.002 MA

Toroidal field, B

0.0875 T

Pulse length

0.0012 s

Ownership

Public

2.40. MT-1 (PAKISTAN ATOMIC ENERGY COMMISSION, PAKISTAN)

2.40.1. Introduction

MT-1 is operating since 2018 by Pakistan Atomic Energy Commission, National Tokamak

Fusion Programme of Pakistan.

2.40.2. Purpose

MT-1 is used for training purposes.

2.40.3. Main features

MT-1 is a small-sized metallic spherical tokamak, with pre-ionization source power up to 3

kW. Technical information is listed in Table 46 below.

TABLE 46. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.25 m

Minor radius, a

0.1 m

Plasma current, I

0.015 MA

Toroidal field, B

1.2 T

Pulse length

0.008 s

Magnetic field configuration

Limiter

Ownership

Public

2.41. MT-2 (PAKISTAN ATOMIC ENERGY COMMISSION, PAKISTAN)

2.41.1. Introduction

MT-2 is the upgraded and elongated version of MT-1 and is currently under construction (coils

system is being manufactured). It is being developed by Pakistan Atomic Energy Commission,

National Tokamak Fusion Programme of Pakistan.

2.41.2. Purpose

MT-2 will be used for training purposes.

2.41.3. Main features

MT-2 vacuum vessel can achieve a vacuum up to ~10

-7

mbars. General information is listed in

Table 47 below.

TABLE 47. GENERAL INFORMATION

Device type

Spherical Tokamak

Status

Under construction

Ownership

Public

2.42. PST (PAKISTAN ATOMIC ENERGY COMMISSION, PAKISTAN)

2.42.1. Introduction

PST tokamak is being developed by Pakistan Atomic Energy Commission, National Tokamak

Fusion Programme of Pakistan. It is currently in the engineering design phase.

2.42.2. Purpose

The objectives of PST are to explore plasma information for steady-state operation of spherical

tokamaks with aspect ratio equal to 2, as well as research and develop high temperature

superconducting coils for tokamak and a liquid lithium divertor system. PST research

programme will also focus on capacity building activities. An extensive training program will

be launched for young scientists and engineers, both nationally and internationally.

2.42.3. Main features

PST will be a medium-sized spherical tokamak planned to have electron Bernstein wave heating

and neutral beam injection systems. A pulsating discharge cleaning system will be used to test

diagnostics [37]. Technical information is listed in Table 48 below.

TABLE 48. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Planned

Major radius, R

0.5 m

Minor radius, a

0.25 m

Plasma current, I

0.31 MA

Toroidal field, B

0.5 T

Ownership

Public

2.43. ISTTOK (INSTITUTO SUPERIOR TÉCNICO, PORTUGAL)

2.43.1. Introduction

ISTTOK is a device for nuclear fusion research and training located at Instituto Superior

Técnico, Portugal.

2.43.2. Purpose

The main objectives of ISTTOK are to: (i) study plasma turbulence; (ii) operation and control

on alternating plasma current regimes; (iii) test liquid metal limiter concepts; (iv) develop and

upgrade plasma-relevant diagnostics for nuclear fusion; and (v) serve as an academic facility to

train master and PhD students.

2.43.3. Main features

ISTTOK is a large aspect ratio circular cross-section iron core tokamak. ISTTOK is highly

suitable for edge physics studies due to its flexibility, low operation costs and short time scale

for diagnostics implementation. Moreover, ISTTOK was the first tokamak equipped with a

switchable insulated-gate bipolar transistor primary circuit with a considerable larger capacitor

bank (3.8 F), allowing multiple alternating current discharges [38]. Technical information is

listed in Table 49 below.

TABLE 49. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.46 m

Minor radius, a

0.085 m

Plasma current, I

0.007 MA

Toroidal field, B

0.8 T

Ownership

Public

2.44. KSTAR (KOREA INSTITUTE OF FUSION ENERGY, REPUBLIC OF KOREA)

2.44.1. Introduction

KSTAR is located at Korean Institute of Fusion Energy, Republic of Korea. Operating since

2008, this device produced over 20000 plasma experimental shots.

2.44.2. Purpose

The main objectives of KSTAR research programme are to research and develop on steady-

state superconducting tokamak physics and establishing a scientific and technological base for

commercial fusion power plants.

2.44.3. Main features

KSTAR is a medium-sized tokamak with superconducting magnets made of Nb3Sn, active

cooled in-vessel components and long-pulse non-inductive heating and current drive. The

device can ensure high performance operational capability thanks to a passive stabilizer, in-

vessel control coils and strong plasma shaping features [39]. Technical information is listed in

Table 50 below.

TABLE 50. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.8 m

Minor radius, a

0.5 m

Plasma current, I

2 MA

Toroidal field, B

3.5 T

Pulse length

300 s

Ownership

Public

2.45. VEST (SEOUL NATIONAL UNIVERSITY, REPUBLIC OF KOREA)

2.45.1. Introduction

Designed and built by Seoul National University, Republic of Korea, VEST is a tokamak

operating since 2013.

2.45.2. Purpose

The main objectives of VEST are to study spherical tokamak plasmas, wave heating and high

beta operations, as well as divertor concepts, including Super-X divertor configuration.

2.45.3. Main features

VEST is a small-sized spherical tokamak. This device can operate in high performance with

high beta thanks to key features [40]. Technical information is listed in Table 51 below.

TABLE 51. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.45 m

Minor radius, a

0.33 m

Plasma current, I

0.1 MA

Toroidal field, B

0.1 T

Magnetic field configuration

Tungsten and graphite inboard limiters

Ownership

Public

2.46. FT-2 (IOFFE INSTITUTE, RUSSIAN FEDERATION)

2.46.1. Introduction

Operating since the 1908s, FT-2 is located at Ioffe Institute, Russian Federation.

2.46.2. Purpose

FT-2 experimental research programme focuses on lower hybrid current drive studies.

2.46.3. Main features

FT-2 is a high aspect ratio conventional tokamak with high toroidal field [41]. Technical

information is listed in Table 52 below.

TABLE 52. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.55 m

Minor radius, a

0.087 m

Plasma current, I

0.04 MA

Toroidal field, B

2.3 T

Magnetic field configuration

Circular limiter

Ownership

Public

2.47. GLOBUS-M2 (IOFFE INSTITUTE, RUSSIAN FEDERATION)

2.47.1. Introduction

Built at Ioffe Institute, Russian Federation, Globus-M2 is the upgraded version of Globus-M,

and it operates since 2018.

2.47.2. Purpose

The main purpose of the Globus-M2 programme is to establish the scientific and technological

basis for compact fusion reactor systems, compact fusion neutron sources and fusion-fission

hybrid systems.

2.47.3. Main features

Globus-M2 is a spherical tokamak, which can operate in both single and double null divertor

configurations. Its electromagnetic system can withstand higher currents and mechanical loads

[42]. Technical information is listed in Table 53 below.

TABLE 53. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.36 m

Minor radius, a

0.24 m

Plasma current, I

0.5 MA

Toroidal field, B

1 T

Magnetic field configuration

Single and double null divertor

2.48. TUMAN-3M (IOFFE INSTITUTE, RUSSIAN FEDERATION)

2.48.1. Introduction

TUMAN-3M is operating at Ioffe Institute, Russian Federation, since 1984.

2.48.2. Purpose

The main purpose of TUMAN-3M is to study plasma confinement and reaction mechanism.

2.48.3. Main features

TUMAN-3M is a tokamak with circular cross-section and without a divertor. The vessel and

circular limiters are made of Inconel, while the sector limiter is made of molybdenum [43].

Technical information is listed in Table 54 below.

TABLE 54. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.53 m

Minor radius, a

0.22 m

Plasma current, I

0.18 MA

Toroidal field, B

1.2 T

Pulse length

0.080 s

Magnetic field configuration

Circular limiter

Ownership

Public

2.49. T-15MD (NATIONAL RESEARCH CENTRE KURCHATOV INSTITUTE,

RUSSIAN FEDERATION)

2.49.1. Introduction

T-15MD is the upgraded version of T-15 tokamak, which operated during 1988–1995 at

Kurchatov Institute, Russian Federation. T-15MD constriction started in 2011 and finished in

2020. The start of operation was celebrated in 2021.

2.49.2. Purpose

T-15MD programme will contribute to ITER’s and future fusion power plants’ operation.

2.49.3. Main features

T-15MD is a medium-sized superconducting tokamak, which will be able to achieve high

plasma temperature and high plasma density, thanks to three auxiliary plasma heating systems

(neutral beam injection, electron cyclotron resonance – i.e., seven gyrotrons and ion cyclotron

resonance heating) and current drive systems (low hybrid heating and current drive with pulse

duration up to 30 s) [44]. Technical information is listed in Table 55 below.

TABLE 55. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

2.43 m

Minor radius, a

0.42 m

Plasma current, I

1 MA

Toroidal field, B

3.5 T

Pulse length

30 s

Magnetic field configuration

Divertor

Ownership

Public

2.50. GUTTA (SAINT PETERSBURG STATE UNIVERSITY, RUSSIAN FEDERATION)

2.50.1. Introduction

GUTTA is located at St. Petersburg State University, Russian Federation, and is operating since

2004.

2.50.2. Purpose

The main objectives of GUTTA are to develop and improve mathematical models applicable

to large tokamaks, study the electron cyclotron resonance heating assisted breakdown and non-

solenoid plasma formation in low aspect ratio tokamak, as well as develop of diagnostics and

train students [45].

2.50.3. Main features

GUTTA is a small spherical tokamak. Technical information is listed in Table 56 below.

TABLE 56. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.16 m

Minor radius, a

0.08 m

Plasma current, I

0.18 MA

Toroidal field, B

1.5 T

Ownership

Public

2.51. T-11M (TROITSK INSTITUTE FOR INNOVATION AND FUSION RESEARCH,

RUSSIAN FEDERATION)

2.51.1. Introduction

Built in 1985, T-11M is located at Troitsk Institute for Innovation and Fusion Research, Russian

Federation.

2.51.2. Purpose

The main purpose of T-11M is to conduct research on lithium plasma facing components

protection and lithium injection into scrape-off-layer plasma.

2.51.3. Main features

T-11M is a small-sized conventional tokamak with a lithium capillary-pore system limiter and

a lithium collecting system in the scrape-off-layer [46]. Technical information is listed in Table

57 below.

TABLE 57. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.7 m

Minor radius, a

0.2 m

Plasma current, I

0.1 MA

Toroidal field, B

1 T

Pulse length

0.25 s

Magnetic field configuration

Capillary-pore system limiter

Ownership

Public

2.52. SMART (UNIVERSITY OF SEVILLE, SPAIN)

2.52.1. Introduction

SMART is being designed at University of Seville, Spain.

2.52.2. Purpose

SMART will serve for educational and training purposes, to study plasma confinement and

stability, plasma facing materials and develop novel technical equipment essential for fusion

research.

2.52.3. Main features

SMART will be a spherical tokamak, which can be operated in three phases depending on

various plasma information [47]. Technical information is listed in Table 58 below.

TABLE 58. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Planned

Major radius, R

0.4 m

Minor radius, a

0.25 m

Plasma current, I

Phase 1 – 0.03 MA

Phase 2 – 0.1 MA

Phase 3 – 0.5 MA

Toroidal field, B

Phase 1 – 0.1 T

Phase 2 – 0.3 T

Phase 3 – 1 T

Pulse length

Phase 1 – 0.002 s

Phase 2 – 0.1 s

Phase 3 – 0.5 s

Magnetic field configuration

Divertor

Ownership

Public

2.53. TCV (SWISS PLASMA CENTER, SWITZERLAND)

2.53.1. Introduction

Operating since 1992, TCV is located at Swiss Federal Institute of Technology Lausanne, Swiss

Plasma Center, Switzerland.

2.53.2. Purpose

The main purpose of TCV is to study plasma configurations and shapes and research magnetic

confinement plasma physics.

2.53.3. Main features

TCV is a medium-sized tokamak with a highly elongated, rectangular vacuum vessel and

sixteen poloidal field coils for plasma formation, equally divided into two stacks located on

both sides of the vessel. An ohmic heating coil drives an inductive current into the plasma [48].

TCV plasma auxiliary heating system consists of neutral beam injector (2 MW), electron

cyclotron resonance heating (3 MW) and electron cyclotron current drive (1.5 MW). Technical

information is listed in Table 59 below.

TABLE 59. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.89 m

Minor radius, a

0.25 m

Plasma current, I

1.2 MA

Toroidal field, B

1.54 T

Pulse length

2.6 s in ohmic, 4 s with electron cyclotron

current drive

Ownership

Public

2.54. TT-1 (THAILAND INSTITUTE OF NUCLEAR TECHNOLOGY, THAILAND)

2.54.1. Introduction

TT-1, known as HT-6M until decommissioned in 2002, is a tokamak donated to Thailand

Institute of Nuclear Technology, Thailand by the Institute of Plasma Physics of the Chinese

Academy of Sciences, China. Plans call for TT-1 to be commissioned by 2023.

2.54.2. Purpose

TT-1 will become the first Thai tokamak and the first fusion device in Southeast Asia. It will

help to develop capacity building and development programmes in the region.

2.54.3. Main features

TT-1 is a small-sized tokamak under construction. In the recommissioning phase, supporting

system of TT-1 (power supply system for the magnet, plasma control system, vacuum system,

data acquisition system) will be redesigned [49]. Technical information is listed in Table 60

below.

TABLE 60. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Under construction

Major radius, R

0.65 m

Minor radius, a

0.2 m

Plasma current, I

0.1 MA

Toroidal field, B

1.52 T

Ownership

Public

2.55. JET (EUROFUSION, UNITED KINGDOM)

2.55.1. Introduction

Operating since 1983, JET is the world's largest tokamak in operation. It is located at Culham

Centre for Fusion Energy, UK. Its scientific programme is run by EUROfusion.

2.55.2. Purpose

The main purpose of JET is to test plasma physics, systems and materials for ITER, and to study

plasma behaviour in conditions and dimensions close to those of a fusion power plant.

2.55.3. Main features

JET is a large tokamak with a divertor installed at the bottom of the vacuum vessel and an

ITER-like first wall made of beryllium and tungsten. JET plasma auxiliary heating system

consists of neutral beam injector (34 MW), ion cyclotron resonance heating (10 MW) and lower

hybrid current drive (7 MW). JET is the only operating tokamak capable of operating with

tritium fuel [50]. Technical information is listed in Table 61 below.

TABLE 61. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

2.96 m

Minor radius, a

1.25 m

Plasma current, I

5 MA

Toroidal field, B

3.4 T

Pulse length

60 s

Magnetic field configuration

Tungsten divertor

Ownership

Public

2.56. ST40 (TOKAMAK ENERGY, UNITED KINGDOM)

2.56.1. Introduction

ST-40 is a fusion device constructed by UK private sector company Tokamak Energy. The

company was founded in 2009.

2.56.2. Purpose

The main purpose of ST-40 is to prove the feasibility of commercial fusion energy with compact

spherical tokamaks.

2.56.3. Main features

ST-40 is a compact spherical tokamak with high temperature superconducting magnets made

of REBCO, manufactured in 0.1 mm thick tapes. ST-40 toroidal field coils are made of copper

[51]. Technical information is listed in Table 62 below.

TABLE 62. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.5 m

Minor radius, a

0.3 m

Plasma current, I

2 MA

Toroidal field, B

3 T

Pulse length

1 s

Magnetic field configuration

Divertor

Ownership

Private

2.57. MAST-U (UKAEA, UNITED KINGDOM)

2.57.1. Introduction

MAST-U is the upgrade of MAST, which operated during 2000–2013. MAST-U’s first plasma

was produced in 2020.

2.57.2. Purpose

The main purpose of MAST-U is to advance plasma exhaust research and support the

development of compact fusion power plants.

2.57.3. Main features

MAST-U is a low aspect ratio tokamak able to operate with a large variety of different divertor

configurations, and first one to operate with Super-X divertor configuration [52]. Technical

information is listed in Table 63 below.

TABLE 63. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.85 m

Minor radius, a

0.65 m

Plasma current, I

2 MA

Toroidal field, B

0.52 T

Pulse length

5 s

Magnetic field configuration

Super-X divertor

Ownership

Public

2.58. HBT-EP (COLUMBIA UNIVERSITY, UNITED STATES OF AMERICA)

2.58.1. Introduction

Operating since 1993, HIBT-EP is located at Columbia University, USA.

2.58.2. Purpose

The main purpose of HBT-EP is to study high-beta tokamak plasma performance.

2.58.3. Main features

HIBT-EP is a conventional tokamak with adjustable walls and a vacuum chamber made with

several quartz cylindrical breaks [53]. Technical information is listed in Table 64 below.

TABLE 64. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

0.94 m

Minor radius, a

0.2 m

Plasma current, I

0.32 MA

Toroidal field, B

0.3 T

Magnetic field configuration

Circular limiter

Ownership

Public

2.59. SPARC (COMMONWEALTH FUSION SYSTEMS, UNITED STATES OF

AMERICA)

2.59.1. Introduction

SPARC is a joint project of US private sector company Commonwealth Fusion Systems (CFS)

and MIT’s Plasma Science and Fusion Center, USA. SPARC is planned to generate a net

scientific energy gain

sci

>1). CFS have disclosed around US$2.1 billion in fusion funding,

which is almost as much as all the other (roughly 30) private sector fusion companies (see Fig.

6 on p.6). SPARC is presently under construction by CFS near Boston, USA.

2.59.2. Purpose

SPARC aims to operate with deuterium and tritium fuel and become the first fusion device that

can produce a net scientific energy gain

2.59.3. Main features

SPARC will be a compact high-field tokamak that will operate with, deuterium and tritium fuel.

It will feature tungsten first walls and high temperature superconducting magnets. In addition

to moderate ohmic heating, SPARC will rely on ion cyclotron resonance heating (25 MW) [54].

Technical information is listed in Table 65 below.

TABLE 65. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Under construction

Major radius, R

1.85 m

Minor radius, a

0.57 m

Plasma current, I

8.7 MA

Toroidal field, B

12.2 T

Ownership

Private

2.60. DIII-D (GENERAL ATOMICS, UNITED STATES OF AMERICA)

2.60.1. Introduction

DIII-D is owned by the US Department of Energy and is located at and managed on their behalf

by General Atomics in San Diego with a large team of collaborating institutions.

2.60.2. Purpose

The main purpose of DIII-D is to develop the plasma solutions and scientific basis to project

them for future fusion reactors. In particular it targets the development of a high performance

fusion core and a compatible power handling solutions to enable a compact fusion power plant,

as well as a range of associated plasma interacting fusion technologies. It also seeks to develop

the operational approach and modelling tools to help ITER maximize performance and reach

its goals rapidly.

2.60.3. Ma

in features

DIII-D’s key feature is its very high degree of flexibility to explore wide range of plasma

configurations and develop solutions for future reactors. This couples with a highly extensive

diagnostic set in order to identify and resolve models of the underlying physics needed to

project to solutions developed. These flexibilities include plasma shapes from highly positive

to negative triangularity, single or double null operation, flexible open and closed divertors,

particle control through cryo-pumping, wall conditioning, gas and inboard or outboard pellet

injection, 3 arrays of 3-D magnetic perturbation coils, and independent control of heating,

current drive, torque and profile breadth [55]. Technical information is listed in Table 66 below.

TABLE 66. TECHNICAL INFORMATION

Device type

Conventional Tokamak

Status

Operating

Major radius, R

1.7 m

Minor radius, a

0.6 m

Plasma current, I

2 MA

Toroidal field, B

2.17 T

Pulse length

10 s

Magnetic field configuration

Divertor, single or double null

Ownership

Public

2.61. LTX-Β (PRINCETON PLASMA PHYSICS LABORATORY, UNITED STATES OF

AMERICA)

2.61.1. Introduction

Operating since 2020, LTX-β is located at Princeton Plasma Physics Laboratory, USA.

2.61.2. Purpose

The purpose of LTX-β is to research and develop plasma facing components based on liquid

lithium.

2.61.3. Main features

LTX-β is a spherical tokamak with low aspect ratio, with a lithium coated shell made of 5 mm

stainless steel bonded to 1 cm thick copper, without use of any low Z materials [56]. Technical

information is listed in Table 67 below.

TABLE 67. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.4 m

Minor radius, a

0.26 m

Plasma current, I

0.175 MA

Toroidal field, B

3.7 T

Ownership

Public

2.62. NSTX-U (PRINCETON PLASMA PHYSICS LABORATORY, UNITED STATES OF

AMERICA)

2.62.1. Introduction

NSTX-U is the upgraded version of NSTX. NSTX-U started operating in 2016 but was paused

to make repairs. It is expected to restart in 2022.

2.62.2. Purpose

The objectives of NSTX-U are: (i) to study plasma control scenarios in a high-performance

spherical tokamak; (ii) to investigate non-inductive plasma smart-up; and (iii) to study heat

exhaust solutions.

2.62.3. Main features

NSTX-U is a spherical tokamak with plasma facing components made of carbon-based

materials. Its vacuum vessel is made of stainless steel. The plasma heating and current drive

systems include the high harmonic fast wave, coaxial helicity injection current drive, the

electron cyclotron resonance and neutral beam injection systems [57]. Technical information is

listed in Table 68 below.

TABLE 68. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.934 m

Minor radius, a

0.68 m

Plasma current, I

2 MA

Toroidal field, B

1 T

Pulse length

5 s

Magnetic field configuration

Divertor

Ownership

Public

2.63. PEGASUS-III (UNIVERSITY OF WISCONSIN-MADISON, UNITED STATES OF

AMERICA)

2.63.1. Introduction

Pegasus is located at University of Wisconsin-Madison, USA.

2.63.2. Purpose

The research purpose of Pegasus is to study non-inductive plasma start up and operational

scenarios.

2.63.3. Main features

Pegasus is a spherical tokamak with stainless steel vacuum vessel [58]. Technical information

is listed in Table 69 below.

TABLE 69. TECHNICAL INFORMATION

Device type

Spherical Tokamak

Status

Operating

Major radius, R

0.45 m

Minor radius, a

0.37 m

Plasma current, I

1 MA

Toroidal field, B

0.58 T

Ownership

Public

3. EXPERIMENTAL STELLARATORS/HELIOTRONS

3.1. CFQS (SOUTHWEST JIAOTONG UNIVERSITY, CHINA)

3.1.1. Introduction

CFQS is a stellarator being constructed as a cooperative project between the National Institute

of Fusion Science, Japan and the Southwest Jiaotong University, China. It will be the first ever

quasi-axisymmetric stellarator.

3.1.2. Purpose

The purpose of CFQS is to research and develop the stellarator concept line.

3.1.3. Main features

CFQS will be a quasi-axisymmetric stellarator. It will feature four poloidal field coils and

twelve toroidal field coils. Its vacuum vessel will be made of SUS316L with thickness of 6 mm

[59]. Technical information is listed in Table 70 below.

TABLE 70. TECHNICAL INFORMATION

Device type

Stellarator

Status

Planned

Major radius, R

1 m

Minor radius, a

0.25 m

Toroidal field, B

1 T

Ownership

Public

3.2. SCR-1 (INSTITUTO TECNOLOGICO DE COSTA RICA, COSTA RICA)

3.2.1. Introduction

Operating since 2016, SCR-1 is a stellarator designed and built by Costa Rica Institute of

Technology, Costa Rica, and the first fusion device in Latin America.

3.2.2. Purpose

The purpose of SCR-1 is to support education and training in plasma physics and fusion

research in Costa Rica and in the Latin America region.

3.2.3. Main features

SCR-1 is the smallest stellarator in the world. It features twelve magnetic coils [60]. Technical

information is listed in Table 71 below.

TABLE 71. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

0.238 m

Minor radius, a

0.042 m

Plasma current, I

0.04 MA

Toroidal field, B

0.0878 T

Ownership

Public

3.3. RENAISSANCE FUSION (RENAISSANCE FUSION, FRANCE)

3.3.1. Introduction

Renaissance fusion is a French private sector company working on stellarators research and

development and developing a stellarator concept [61].

3.3.2. Purpose

The purpose of Renaissance fusion is to put fusion electricity in the grid using the stellarator

concept.

3.3.3. Main features

Renaissance Fusion’s stellarator design is intended to feature high temperature superconducting

magnets and operate with flowing liquid metal walls for the plasma facing components.

Technical information is listed in Table 72 below.

TABLE 72. GENERAL INFORMATION

Device type

Stellarator

Status

Planned

Toroidal field, B

10 T

Ownership

Private

3.4. WENDELSTEIN 7-X (MAX PLANK INSTITUTE FOR PLASMA PHYSICS,

GERMANY)

3.4.1. Introduction

Wendelstein 7-X is located at Max Planck Institute in Greifswald, Germany. Operating since

2015, it is the largest stellarator device in the world.

3.4.2. Purpose

The main purpose of Wendelstein 7-X is to prove the possibility of using the stellarator concept

for power plant designs, by operating up to 30 minutes with a heating power up to 10 MW [62].

3.4.3. Main features

Wendelstein 7-X is a large stellarator type fusion device with modular superconducting coils.

The vessel needs to accurately repeat the swirling shape of the plasma and it is made of stainless

steel segments 17 mm thick, accurately bent and welded to each other. The magnetic system of

Wendelstein 7-X includes 20 planar and 50 non-planar superconducting magnetic coils. Since

2022 all plasma facing components are water-cooled. Technical information is listed in Table

73 below.

TABLE 73. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

5.5 m

Minor radius, a

0.53 m

Toroidal field, B

3 T

Ownership

Public

3.5. TJ-K (UNIVERSITY OF STUTTGART, GERMANY)

3.5.1. Introduction

TJ-K is a stellarator which was built by CIEMAT, Madrid and it is located at the University of

Stuttgart, Germany, since 2005.

3.5.2. Purpose

The main purpose of TJ-K is to carry out low temperature plasmas research.

3.5.3. Main features

TJ-K consists of two vertical field coils and one helical coil, which is wrapped six times around

the vacuum vessel. These three coils are generating toroidally closed magnetic flux surfaces.

TJ-K operates with low temperature plasmas [63]. Technical information is listed in Table 74

below.

TABLE 74. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

0.6 m

Minor radius, a

0.1 m

Toroidal field, B

0.07 T

Ownership

Public

3.6. HELIOTRON J (KYOTO UNIVERSITY, JAPAN)

3.6.1. Introduction

Heliotron J was designed and manufactured by Kyoto University, Japan, in 2000.

3.6.2. Purpose

The main objectives of Heliotron J are to study the concept of the non-symmetric quasi-

isodynamic approach to the heliotron line of stellarators optimization with a helical axis, as well

as to investigate the physics basis for the heliotron helical axis.

3.6.3. Main features

Heliotron J is a stellarator with a three-dimensional magnetic axis, so called helical magnetic

axis. The magnetic coil system includes a continuous helical coil, two types of toroidal coils

and three pairs of vertical coils. The heating system consists of electron cyclotron resonance

heating (70 GHz, 0.4 MW), 2 neutral beam injectors (30 kV, 0.7MW) and 2 ion cyclotron

resonance heating (16–24 MHz, 0.4 MW) [64]. Technical information is listed in Table 75

below.

TABLE 75. TECHNICAL INFORMATION

Device type

Heliotron

Status

Operating

Major radius, R

1.2 m

Minor radius, a

0.17 m

Toroidal field, B

1.5 T

Ownership

Public

3.7. LHD (NATIONAL INSTITUTE FOR FUSION SCIENCE, JAPAN)

3.7.1. Introduction

Operating since 1998, the LHD fusion device is one of the largest stellarators in the world. LHD

is located at the National Institute for Fusion Science, Japan.

3.7.2. Purpose

The main purpose of LHD is to contribute to research on the development of helical-type fusion

devices.

3.7.3. Main features

LHD is a large heliotron type device. It operates using superconducting coils, various plasma

heating systems and multiple diagnostics [65]. Technical information is listed in Table 76

below.

TABLE 76. TECHNICAL INFORMATION

Device type

Heliotron

Status

Operating

Major radius, R

3.9 m

Minor radius, a

0.65 m

Toroidal field, B

4 T

Ownership

Public

3.8. TJ-II (CIEMAT, SPAIN)

3.8.1. Introduction

Operating since 1997, TJ-II is a joint project between the National Fusion Laboratory, Spain

and Oak Ridge National Laboratory, USA. It was also partly financed by EURATOM. After

Wendelstein 7-X, TJ-II is the second largest stellarator in Europe.

3.8.2. Purpose

The main purpose of TJ-II is to study magnetic configuration influence on heat and particle

transport.

3.8.3. Main features

TJ-II is a flexible, medium-sized stellarator. It operates with 32 toroidal field coils, one circular

coil, one helical coil and a set of vertical field coils. Power heating is provided by means of

microwave heating (800 kW, 53 GHz) and neutral beam injector (1.6 MW, 30 keV) [66].

Technical information is listed in Table 77 below.

TABLE 77. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

1.5 m

Minor radius, a

0.22 m

Toroidal field, B

0.95 T

Ownership

Public

3.9. URAGAN-2M (INSTITUTE OF PLASMA PHYSICS NATIONAL SCIENCE

CENTER, UKRAINE)

3.9.1. Introduction

Operating since 2006, Uragan-2M is a located at Institute of Plasma Physics National Science

Center in Kharkiv Institute of Physics and Technology, Ukraine.

3.9.2. Purpose

The main purpose of Uragan-2M is to conduct studies on the influence of helical magnetic field

inhomogeneities in plasma confinement.

3.9.3. Main features

Uragan-2M is a stellarator. Its configuration with reduced helical magnetic can moderate sheers

and magnetic wells. Uragan-2M features with 16 toroidal field coils and 2 helical coils. The

large number of magnetic windings makes it possible to vary the information of the magnetic

configuration over a wide range. Technical information is listed in Table 78 below.

TABLE 78. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

1.7 m

Minor radius, a

0.22 m

Toroidal field, B

2.4 T

Ownership

Public

3.10. URAGAN-3M (INSTITUTE OF PLASMA PHYSICS NATIONAL SCIENCE

CENTER, UKRAINE)

3.10.1. Introduction

Uragan-3M is located at Institute of Plasma Physics National Science Center in Kharkiv

Institute of Physics and Technology, Ukraine.

3.10.2. Purpose

The main purpose of Uragan-3M is to conduct research on radiofrequency plasma production

and heating, stellarator plasma physics and divertor study in the stellarator-line research.

3.10.3. Main features

Uragan-3M is a medium-sized stellarator. Its magnetic system is located inside the vacuum

chamber, allowing for helical divertor configuration. Technical information is listed in Table

79 below.

TABLE 79. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

1 m

Minor radius, a

0.12 m

Toroidal field, B

0.7 T

Ownership

Public

3.11. CTH (AUBURN UNIVERSITY, UNITED STATES OF AMERICA)

3.11.1. Introduction

Operating since 2005, CTH is located at Auburn University, USA.

3.11.2. Purpose

The main purpose of CTH is to investigate magnetohydrodynamics instabilities and disruptions

in conductive stellarator plasmas.

3.11.3. Main features

CTH is a stellarator of torsatron type. Its magnetic system consists of external helical, vertical,

and toroidal field coils. In addition, CTH can also operate with a set of ohmic coils typical of

pulsed tokamak designs. The vacuum vessel is made of Inconel alloy 625 and has a circular

torus shape [67]. Technical information is listed in Table 80 below.

TABLE 80. TECHNICAL INFORMATION

Device type

Torsatron

Status

Operating

Major radius, R

0.75 m

Minor radius, a

0.29 m

Plasma current, I

0.08 MA

Toroidal field, B

0.7 T

Ownership

Public

3.12. HIDRA (UNIVERSITY OF ILLINOIS, UNITED STATES OF AMERICA)

3.12.1. Introduction

Formerly known as WEGA, HIDRA is located at University of Illinois, USA. It operates since

2016.

3.12.2. Purpose

The main purpose of HIDRA is to study plasma material interactions and develop the

technology essential for innovative plasma facing components (e.g., using liquid lithium

technologies).

3.12.3. Main features

HIDRA is a medium-sized device which can operate either as a stellarator or tokamak. Its

vacuum vessel has a circular shape. Plasma in HIDRA is generated using magnetron heating

(26 kW, 2.45 GHz). Technical information is listed in Table 81 below.

TABLE 81. TECHNICAL INFORMATION

Device type

Stellarator/Tokamak

Status

Operating

Major radius, R

0.72 m

Minor radius, a

0.19 m

Toroidal field, B

0.5 T

Ownership

Public

3.13. HSX (UNIVERSITY OF WISCONSIN-MADISON, UNITED STATES OF

AMERICA)

3.13.1. Introduction

Operating since 1999, HSX is located at University of Wisconsin-Madison, USA. HSX has a

unique structure of the magnetic field, called the Quasi-Helically Symmetric (QHS).

3.13.2. Purpose

The main purpose of HSX is to conduct research in investigation of transport, turbulence and

confinement in a QHS magnetic field.

3.13.3. Main features

The QHS configuration of HSX is achieved thanks to 48 magnetic coils and a set of 12 auxiliary

coils [100]. HSX vacuum vessel is made of stainless steel. Plasma is generated using gyrotron

system (200 kW, 28 GHz) [68]. Technical information is listed in Table 82 below.

TABLE 82. TECHNICAL INFORMATION

Device type

Stellarator

Status

Operating

Major radius, R

1.2 m

Minor radius, a

0.15 m

Toroidal field, B

1.25 T

Ownership

Public

4. EXPERIMENTAL INERTIAL/LASER FUSION DEVICES

4.1. HB11 (HB11 ENERGY, AUSTRALIA)

4.1.1. Introduction

HB11 is a laser fusion device being developed by HB11 Energy, an Australian private sector

company launched in 2019 [61].

4.1.2. Purpose

The purpose is to show the economic advantage of p-

B fusion reaction for energy production.

4.1.3. Main features

The laser system will be composed of a nanosecond pulse laser and a picosecond pulse laser.

The device will produce energy via p-

B fusion reactions. General information is listed in

Table 83 below.

TABLE 83. GENERAL INFORMATION

Device type

Laser Fusion

Status

Planned

Ownership

Private

4.2. LMJ (CEA, FRANCE)

4.2.1. Introduction

LMJ is located at CEA, France. It was commissioned at the end of 2014. Since 2017, it has

been merged with the high-power PETAL laser.

4.2.2. Purpose

The main purpose of LMJ is to support high energy density physics studies and inertial

confinement fusion experiments.

4.2.3. Main features

LMJ is designed with multiple lines of a three-frequency neodymium glass laser beam capable

of irradiating targets at a wavelength of 351 nm [69]. General information is listed in Table 84

below.

TABLE 84. GENERAL INFORMATION

Device type

Laser Fusion

Status

Operating

Ownership

Public

4.3. MARVEL FUSION (MARVEL FUSION, GERMANY)

4.3.1. Introduction

Established in 2018, Marvel Fusion is a German private sector company planning to develop

laser fusion device technology [61].

4.3.2. Purpose

The main purpose of Marvel Fusion is to show the feasibility of commercial laser fusion via p-

B reactions.

4.3.3. Main features

Marvel Fusion concept is based on ultra-short pulses, high intensity lasers and manufactured

nanostructured fuel pellets triggering p-

B fusion reactions. General information is listed in

Table 85 below.

TABLE 85. GENERAL INFORMATION

Device type

Laser Fusion

Status

Planned

Ownership

Private

100

4.4. GEKKO XII (OSAKA UNIVERSITY, JAPAN)

4.4.1. Introduction

Built in 1983 and located at Osaka University, Japan, the GEKKO XII is a large-scale laser

facility capable of producing high temperature plasma (hundred million degrees Celsius) and

compressing matter to density greater than 600 times its solid density.

4.4.2. Purpose

The main purpose of GEKKO XII is to support high energy density physics studies and inertial

confinement fusion experiments.

4.4.3. Main features

GEKKO XII is a large scale powerful 12-beam glass laser. Its output energy is 20 kJ, and its

peak power is 40 TW. The GEKKO XII laser operates with two vacuum chambers. In chamber,

the target can be irradiated by 12 laser beams with spherical symmetry, while in another

chamber, the 12 beams are combined into a single laser beam which allows to irradiate the

target with high-power from a single direction [70]. General information is listed in Table 86

below.

TABLE 86. GENERAL INFORMATION

Device type

Laser Fusion

Status

Operating

Ownership

Public

101

4.5. LFEX (OSAKA UNIVERSITY, JAPAN)

4.5.1. Introduction

LFEX is a laser fusion facility at Osaka University, Japan.

4.5.2. Purpose

The main purpose of LFEX is to conduct fusion research and study relativistic plasma

interactions and laser-induced nuclear physics.

4.5.3. Main features

LFEX is an ultra-high intensity laser facility designed to produce an output energy of 10 kJ

[71]. General information is listed in Table 87 below.

TABLE 87. GENERAL INFORMATION

Device type

Laser Fusion

Status

Operating

Ownership

Public

102

4.6. FIRST LIGHT (FIRST LIGHT FUSION LTD, UNITED KINGDOM)

4.6.1. Introduction

First Light Fusion Ltd is a UK private sector company developing inertial fusion technology

[61].

4.6.2. Purpose

First Light Fusion purpose is to achieve energy generation by inertial confinement fusion using

shockwaves.

4.6.3. Main features

In First Light Fusion’s device, high temperatures and compressions required for fusion are

achieved using a gas gun that launches a projectile into a vacuum chamber at speeds of more

than 6.5 km/s. General information is listed in Table 88 below.

TABLE 88. GENERAL INFORMATION

Device type

Inertial Fusion

Status

Operating

Ownership

Private

103

4.7. INNOVEN ENERGY LLC (INNOVEN ENERGY, UNITED STATES OF AMERICA)

4.7.1. Introduction

Innoven Energy is a US private sector company founded in 2010 and working on laser fusion

R&D.

4.7.2. Purpose

The purpose of Innoven Energy is to show the possibility of producing controlled and safe

inertial fusion energy.

4.7.3. Main features

The Innoven Energy’s device is planned to be based on a novel laser architecture that allows

the nanosecond time compression of laser pulses with extremely high optical quality at low

cost. General information is listed in Table 89 below.

TABLE 89. GENERAL INFORMATION

Device type

Laser Fusion

Status

Planned

Ownership

Private

104

4.8. NIF (LAWRENCE LIVERMORE NATIONAL LABORATORY, UNITED STATES OF

AMERICA)

4.8.1. Introduction

Operating since 2009 and located at Lawrence Livermore National Laboratory, USA, NIF is

the largest laser facility in the world.

4.8.2. Purpose

The main purpose of NIF’s lasers includes laser fusion R&D, high energy density science,

energy security, and building future generations of scientists.

4.8.3. Main features

NIF is the largest and highest energy laser system in the world. NIF consists of 192 high energy,

finely focused laser beams that converge at the centre of the target chamber [72]. General

information is listed in Table 90 below.

TABLE 90. GENERAL INFORMATION

Device type

Laser Fusion

Status

Operating

Ownership

Public

105

4.9. OMEGA (UNIVERSITY OF ROCHESTER LABORATORY FOR LASER

ENERGETICS, UNITED STATES OF AMERICA)

4.9.1. Introduction

Operating since 1995, the OMEGA laser is located at University of Rochester, USA.

4.9.2. Purpose

The main purpose of OMEGA is to advance high energy density science and conduct research

on the interaction of a laser with a substance of ultrahigh intensity.

4.9.3. Main features

OMEGA consists of 60 laser beams capable of focusing up to 30 kJ of energy on a target smaller

than 1 mm in diameter in about one billionth of a second [73]. General information is listed in

Table 91 below.

TABLE 91. GENERAL INFORMATION

Device type

Laser Fusion

Status

Operating

Ownership

Public

107

5. EXPERIMENTAL ALTERNATIVE DEVICE CONCEPTS

5.1. GENERAL FUSION (GENERAL FUSION INC, CANADA)

5.1.1. Introduction

Established in 2002, General Fusion is a Canadian private company developing an experimental

fusion device based on magnetized target fusion [61].

5.1.2. Purpose

The main purpose of General Fusion is to research and develop magnetized target fusion.

5.1.3. Main features

General Fusion developing magnetized target fusion based on the use of liquid metal wall in

combination with lithium for tritium breeding. General information is listed in Table 92 below.

TABLE 92. GENERAL INFORMATION

Device type

Magnetized Target Fusion

Status

Under construction

Ownership

Private

108

5.2. KTX (UNIVERSITY OF SCIENCE AND TECHNOLOGY OF CHINA, CHINA)

5.2.1. Introduction

KTX is a reversed field pinch device located at University of Science and Technology of China.

5.2.2. Purpose

The main objective of KTS is to study resistive wall instabilities.

5.2.3. Main features

KTX is a middle-sized fusion device of reversed field pinch type. It consists of the vacuum

vessel, the conducting shell, ohmic heating, plasma equilibrium, toroidal field and active

feedback coils and the supporting structures [74]. Technical information is listed in Table 93

below.

TABLE 93. TECHNICAL INFORMATION

Device type

Reversed Field Pinch

Status

Operating

Major radius, R

1.4 m

Minor radius, a

0.4 m

Plasma current, I

1 MA

Ownership

Public

109

5.3. TORIX (ÉCOLE POLYTECHNIQUE, FRANCE)

5.3.1. Introduction

ToriX is located at Ecole Polytechnique, France.

5.3.2. Purpose

The purpose of ToriX is to support training and education in fusion research and plasma

physics.

5.3.3. Main features

ToriX is a small-sized device of simple magnetized torus type. Technical information is listed

in Table 94 below.

TABLE 94. TECHNICAL INFORMATION

Device type

Simple Magnetized Torus

Status

Operating

Major radius, R

0.6 m

Minor radius, a

0.05 m

Pulse length

0.1 s

Ownership

Public

110

5.4. RFX (CONSORZIO RFX, ITALY)

5.4.1. Introduction

Opiating since 1992, RFX is the largest reversed field pinch device in the world.

5.4.2. Purpose

The main objectives of RFX are to research and develop the reversed field pinch configuration

and support ITER research plan.

5.4.3. Main features

RFX is a reversed field pinch device featuring ohmic plasma heating [75]. Technical

information is listed in Table 95 below.

TABLE 95. TECHNICAL INFORMATION

Device type

Reversed Field Pinch

Status

Operating

Major radius, R

2 m

Minor radius, a

0.45 m

Plasma current, I

2 MA

Ownership

Public

111

5.5. RELAX (KYOTO INSTITUTE OF TECHNOLOGY, JAPAN)

5.5.1. Introduction

RELAX is a reversed field pinch fusion device located at the Kyoto Institute of Technology,

Japan.

5.5.2. Purpose

The main purpose of RELAX is to research and develop the reversed field pinch in the low

aspect ratio regime.

5.5.3. Main features

RELAX is a low aspect ratio reversed field pinch device [76]. Technical information is listed

in Table 96 below.

TABLE 96. TECHNICAL INFORMATION

Device type

Reversed Field Pinch

Status

Operating

Major radius, R

0.5 m

Minor radius, a

0.25 m

Ownership

Public

112

5.6. UH-CTI (KYUSHU UNIVERSITY, JAPAN)

5.6.1. Introduction

Operating since 2005, UH-CTI is located at Kyusyu University, Japan. Since 2012, UH-CTI is

installed on QUEST tokamak (see p. 47).

5.6.2. Purpose

The main purpose of UH-CTI is to study advanced fuelling in spherical tokamak plasmas.

5.6.3. Main features

UH-CTI is a compact torus used a particles injector and installed on QUEST tokamak. UH-CTI

is mounted perpendicular to the magnetic axis on the midplane of QUEST [77]. General

information is listed in Table 97 below.

TABLE 97. GENERAL INFORMATION

Device type

Spheromak

Status

Operating

Ownership

Public

113

5.7. FAT-CM (NIHON UNIVERSITY, JAPAN)

5.7.1. Introduction

FAT-CM is located at Nihon University, Japan.

5.7.2. Purpose

The purpose of FAT-CM is to investigate the collisional-merging formation process in field

reversed configuration devices at super Alfvén velocity.

5.7.3. Main features

FAT-CM consists of the central confinement chamber and two field reversed theta-pinch

formation sections. The formation tubes are made of transparent quartz and the confinement

chamber is made of stainless steel. Initial field reversed configurations are formed with D

gas

puffing [78]. General information is listed in Table 98 below.

TABLE 98. GENERAL INFORMATION

Device type

Field Reversed Configuration

Status

Operating

Ownership

Private

114

5.8. RT-1 (THE UNIVERSITY OF TOKYO, JAPAN)

5.8.1. Introduction

Operating since 2006, RT-1 is located at University of Tokyo, Japan.

5.8.2. Purpose

The main purpose of RT-1 is to demonstrate very high-beta (~1) plasma confinement.

5.8.3. Main features

RT-1 is a levitated dipole device equipped with a superconducting ring magnet that generates

a dipole magnetic field with strength ranging 0.01-0.3 T. The superconducting ring is levitated

in the middle of the chamber by a feedback-controlled magnet placed on the top of the device

[79]. General information is listed in Table 99 below.

TABLE 99. GENERAL INFORMATION

Device type

Levitated Dipole

Status

Operating

Ownership

Public

115

5.9. UH-MCPG1 (UNIVERSITY OF HYOGO, JAPAN)

5.9.1. Introduction

UH-MCPG1 is located at University of Hyogo, Japan.

5.9.2. Purpose

The purpose of UH-MCPG1 is to carry out research relevant to fusion and plasma physics.

5.9.3. Main features

UH-MCPG1 is a spheromak. General information is listed in Table 100 below.

TABLE 100. GENERAL INFORMATION

Device type

Spheromak

Status

Operating

Ownership

Public

116

5.10. GAMMA 10/PDX (UNIVERSITY OF TSUKUBA, JAPAN)

5.10.1. Introduction

Operating since 1983, GAMMA 10/PDX is located at University of Tsukuba, Japan.

5.10.2. Purpose

The main purpose of GAMMA 10/PDX is to study divertor physics, using high heat-flux and

particle flux plasma.

5.10.3. Main features

GAMMA 10/PDX is a magnetic mirror machine and the largest tandem mirror in the world.

Its plasma heating system consists of radiofrequency wave and neutral beam injection [80].

General information is listed in Table 101 below.

TABLE 101. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Operating

Ownership

Public

117

5.11. PILOT GAMMA PDX-SC (UNIVERSITY OF TSUKUBA, JAPAN)

5.11.1. Introduction

Pilot GAMMA PDX-SC is being constructed at University of Tsukuba, Japan. The installation

of two superconducting coils was completed in 2021.

5.11.2. Purpose

The purpose of Pilot GAMMA PDX-SC is to serve as a pilot device for building a database of

results, contributing to the construction of a future divertor plasma generator.

5.11.3. Main features

Pilot GAMMA PDX-SC will be a magnetic mirror machine with central plasma diameter

ranging 0.4–1.1 m, mirror diameter of 0.1 m, plasma diameter of 0.2 m, magnetic field strength

on the central axis ranging 0.05–0.1 T, and mirror end magnetic field strength of 1.5 T [80].

General information is listed in Table 102 below.

TABLE 102. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Under construction

Ownership

Public

118

5.12. CAT (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN FEDERATION)

5.12.1. Introduction

CAT is being constructed at Budker Institute of Nuclear Physics, Russian Federation.

5.12.2. Purpose

The main purpose of CAT is to study production and sustainment of high pressure plasmas.

5.12.3. Main features

CAT is an axisymmetric magnetic mirror machine. CAT will consist of a central chamber, a

plasma gun and a plasma ejection chamber with a target for absorbing plasma flowing from a

mirror trap. It will produce fast ions by injecting neutral beams (3.5 MW, 15 keV, 5 ms) into a

compact axisymmetric mirror trap [81]. General information is listed in Table 103 below.

TABLE 103. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Under construction

Ownership

Public

119

5.13. GDMT (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN

FEDERATION)

5.13.1. Introduction

GDMT is being designed and planned at Budker Institute of Nuclear Physics, Russian

Federation. Its construction is expected to be completed by 2024.

5.13.2. Purpose

The main purpose of GDMT will be to evaluate and prove the feasibility of the gas dynamic

mirror concept for different practical fusion applications.

5.13.3. Main features

GDMT is being designed as an axisymmetric magnetic mirror trap. It will produce pulses of

5 s duration [82]. General information is listed in Table 104 below.

TABLE 104. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Planned

Ownership

Public

120

5.14. GDMT CORE (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN

FEDERATION)

5.14.1. Introduction

GDMT CORE is being designed and planned at Budker Institute of Nuclear Physics, Russian

Federation. It will be the upgraded and final version of GDMT (see previous page).

5.14.2. Purpose

The main purpose of GDMT CORE will be to evaluate and prove the feasibility of the gas

dynamic mirror concept for different practical fusion applications.

5.14.3. Main features

GDMT CORE is being designed as an axisymmetric magnetic mirror trap. GDMT CORE will

be 70 m in length and produce pulses of 1000 s duration. The superconducting magnet system

of GDMT CORE will consist of a central solenoid, compact mirror coils with high magnetic

field and tail sections to suppress plasma flux [82]. General information is listed in Table 105

below.

TABLE 105. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Planned

Ownership

Public

121

5.15. GDT (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN FEDERATION)

5.15.1. Introduction

Built in 1986, GDT operates at Budker Institute of Nuclear Physics, Russian Federation.

5.15.2. Purpose

The main purpose of GDT is to study magnetic mirror machine physics.

5.15.3. Main features

GDT is a magnetic mirror machine. It is a long axial-symmetric mirror system (with mirror

ratio ranging 12.5–100) that confines fast ions (produced with neutral beam injection system)

and a collisional target plasma. General information is listed in Table 106 below.

TABLE 106. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Operating

Ownership

Public

122

5.16. GOL-NB (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN

FEDERATION)

5.16.1. Introduction

GOL-NB is located at the Budker Institute of Nuclear Physics, Russian Federation. It serves for

testing the main components of the larger GDMT project (see pp. 119–120).

5.16.2. Purpose

The main purpose of the GOL-NB device is to support GDMT’s design and construction.

5.16.3. Main features

GOL-NB is an operational magnetic mirror machine. Its magnetic system consists of a central

trap for plasma confinement and two multiply mirrors for energy and particle confinement.

GOL-NB can work as a classical gas dynamic trap (with short mirrors), or a trap with long

collisional mirrors, as well as with multiple mirrors system. The plasma heating system consists

of two neutral beam injectors (25 keV, 0.75 MW) [83]. General information is listed in Table

107 below.

TABLE 107. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Operating

Ownership

Public

123

5.17. SMOLA (BUDKER INSTITUTE OF NUCLEAR PHYSICS, RUSSIAN

FEDERATION)

5.17.1. Introduction

SMOLA is a magnetic mirror machine located at Budker Institute of Nuclear Physics, Russian

Federation.

5.17.2. Purpose

The main purpose of SMOLA is to explore plasma flow suppression in a helical magnetic field.

5.17.3. Main features

SMOLA is a magnetic mirror machine. It consists of a plasma tank, a solenoid with independent

windings for uniform and helical field components, a set of bias electrodes and an exit tank

with plasma receiver [84]. General information is listed in Table 108 below.

TABLE 108. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Operating

Ownership

Public

124

5.18. EXTRAP T2R (KTH ROYAL INSTITUTE OF TECHNOLOGY, SWEDEN)

5.18.1. Introduction

Operating since 1994, EXTRAP T2R is located at Royal Institute of Technology in Stockholm,

Sweden.

5.18.2. Purpose

The main objective of EXTRAP T2R is to study the stability of resistive wall modes.

5.18.3. Main features

EXTRAP T2R is a medium-sized fusion device of reversed field pinch type. Its ring-shaped

plasma chamber features various access ports for insertion of material samples and probes.

General information is listed in Table 109 below.

TABLE 109. GENERAL INFORMATION

Device type

Reversed Field Pinch

Status

Operating

Ownership

Public

125

5.19. TORPEX (SWISS PLASMA CENTER, SWITZERLAND)

5.19.1. Introduction

TORPEX is located at Center for Plasma Physics Research, Switzerland.

5.19.2. Purpose

The main purpose of TORPEX is to carry out research in plasma physics [85].

5.19.3. Main features

TORPEX is a simple magnetized torus featuring 28 toroidal magnetic coils, 4 vertical coils and

a central coil stack for high toroidal loop voltage operations. Technical information is listed in

Table 110 below.

TABLE 110. TECHNICAL INFORMATION

Device type

Simple Magnetized Torus

Status

Operating

Major radius, R

1 m

Minor radius, a

0.2 m

Plasma current, I

0.001 MA

Ownership

Public

126

5.20. FUSION POWER CORE (COMPACT FUSION SYSTEMS, UNITED STATES OF

AMERICA)

5.20.1. Introduction

Fusion Power Core is a device being planned by US private company Compact Fusion Systems

[61].

5.20.2. Purpose

The main purpose of Fusion Power Core is to research and develop magnetized target fusion.

5.20.3. Main features

Fusion Power Core is being designed as a magnetized target fusion device. General information

is listed in Table 111 below.

TABLE 111. GENERAL INFORMATION

Device type

Magnetized Target Fusion

Status

Planned

Ownership

Private

127

5.21. IDCD (CTFUSION, UNITED STATES OF AMERICA)

5.21.1. Introduction

IDCD is a device developed and operated by US private company CTFusion [61].

5.21.2. Purpose

The purpose of IDCD purpose is to demonstrate reactor-relevant plasma operation.

5.21.3. Main features

IDCD is based on an approach called dynomak, in which plasmas are formed and maintained

by non-axisymmetric, fully inductive magnetic helicity injectors. General information is listed

in Table 112 below.

TABLE 112. GENERAL INFORMATION

Device type

Spheromak

Status

Operating

Ownership

Private

128

5.22. HELICITY DRIVE (HELICITYSPACE, UNITED STATES OF AMERICA)

5.22.1. Introduction

Helicity Drive is a device being planned by US private company Helicity Space [61].

5.22.2. Purpose

The main purpose of Helicity Drive is to make space missions faster and more efficient based

on fusion-driven power and propulsion systems.

5.22.3. Main features

Helicity Drive is being designed as a device for both space propulsion and power. Its systems

are based on plasma magnetic reconnection and magnetic compression with passive coils.

General information is listed in Table 113 below.

TABLE 113. GENERAL INFORMATION

Device type

Space Propulsor

Status

Planned

Ownership

Private

129

5.23. POLARIS (HELION ENERGY, UNITED STATES OF AMERICA)

5.23.1. Introduction

Polaris is a device being planned by US company Helion Energy. Its construction is expected

to start in 2023 [61]. Helion Energy is designing a pulsed fusion energy generator based on self-

organized, high beta plasmas in a field reversed configuration. The high beta of the plasma

would enable direct conversion of the fusion energy held in the charged particles in the plasma

into electricity, which is efficient as compared to thermal cycle conversion of neutron energy.

Such a design motivates the use of alternate fuels like D-

He (see Table 1, p. 4), where most of

the fusion energy is released in charged particles as opposed to the D-T case, in which most of

the fusion energy is released in neutrons.

5.23.2. Purpose

The purpose of Polaris is to achieve

He production via D-D fusion reactions (see Table 1, p.

4).

5.23.3. Main features

Polaris is being designed as a field reversed configuration device. General information is listed

in Table 114 below.

TABLE 114. GENERAL INFORMATION

Device type

Field Reversed Configuration

Status

Planned

Ownership

Private

130

5.24. TRENTA (HELION ENERGY, UNITED STATES OF AMERICA)

5.24.1. Introduction

Trenta is a device developed and operated by US private company Helion Energy. Its

construction was completed in 2020 [61].

5.24.2. Purpose

The purpose of Trenta is to support Helion Energy’s research and development programme,

with the aim of demonstrating fusion conditions and achieving

He production via D-D fusion

reactions (see Table 1, p. 4).

5.24.3. Main features

Trenta is a field reversed configuration device. Technical information is listed in Table 115

below.

TABLE 115. TECHNICAL INFORMATION

Device type

Field Reversed Configuration

Status

Operating

Magnetic field, B

>10 T

Plasma lifetime

0.001 s

Ownership

Private

131

5.25. HORNE HYBRID REACTOR (HORNE TECHNOLOGIES LLC, UNITED STATES

OF AMERICA)

5.25.1. Introduction

Horne Hybrid Reactor is a device designed and being constructed by US private company

Horne Technologies LLC [61].

5.25.2. Purpose

The main purpose of Horne Hybrid Reactor is to optimize plasma confinement and energy

balance in this class of inertial electrostatic fusion devices.

5.25.3. Main features

Horne Hybrid Reactor is an inertial electrostatic fusion device under construction. It will use

REBCO superconducting magnets for producing a high-beta magnetic configuration. The

device will rely on inertial electrostatic confinement for plasma heating. General information

is listed in Table 116 below.

TABLE 116. GENERAL INFORMATION

Device type

Inertial Electrostatic Fusion

Status

Under construction

Ownership

Private

132

5.26. PJMIF (HYPERJET FUSION CORPORATION, UNITED STATES OF AMERICA)

5.26.1. Introduction

PJMIF is a device being planned by US private company HyperJet Fusion Corporation, whose

focus is designing and building plasma guns [61].

5.26.2. Purpose

The purpose of PJMIF is to study the feasibility of driving pulsed fusion energy reaction via

imploding plasma liners.

5.26.3. Main features

PJMIF is being designed as a magnetized target fusion device in which plasma is formed at the

centre of a spherical vacuum vessel. Fusion combustion is achieved by compressing the plasma

to a diameter of 1 cm, thanks to the ejection of plasma jets with hypersonic velocities from the

periphery of the vacuum vessel. The jets merge forming a plasma liner, which continues to

narrow towards the centre. The device will feature a liquid wall to absorb the fusion neutrons

from D-T reactions (see Table 1, p. 4), breeding tritium and eventually serving as a coolant in

a heat exchanger. General information is listed in Table 117 below.

TABLE 117. GENERAL INFORMATION

Device type

Magnetized Target Fusion

Status

Planned

Ownership

Private

133

5.27. FOCUS FUSION (LAWRENCEVILLE PLASMA PHYSICS, INC. DBA

LPPFUSION, UNITED STATES OF AMERICA)

5.27.1. Introduction

Focus Fusion is a device located at Lawrenceville Plasma Physics, Inc., USA [61].

5.27.2. Purpose

The purpose of Focus Fusion is to pave the way for net fusion energy from p-

B reactions (see

Table 1, p. 4).

5.27.3. Main features

Focus Fusion is a dense plasma focus designed to work with p-

B fuel. It consists of two

cylindrical metal electrodes, nested inside each other, and placed inside the vacuum chamber.

General information is listed in Table 118 below.

TABLE 118. GENERAL INFORMATION

Device type

Dense Plasma Focus

Status

Operating

Ownership

Private

134

5.28. CFR (LOCKHEED MARTIN, UNITED STATES OF AMERICA)

5.28.1. Introduction

CFR is a device designed and operated by Lockheed Martin. The project started in 2010.

5.28.2. Purpose

The purpose of CFR is to demonstrate high beta plasma operation in a compact magnetic mirror

machine.

5.28.3. Main features

CFR is a compact magnetic mirror machine able to achieve high beta plasma by combining

cusp confinement and magnetic mirrors. CFR features superconducting magnets. General

information is listed in Table 119 below.

TABLE 119. GENERAL INFORMATION

Device type

Magnetic Mirror Machine

Status

Operating

Ownership

Public

135

5.29. MIFTI (MAGNETO-INERTIAL FUSION TECHNOLOGIES, INC., UNITED

STATES OF AMERICA)

5.29.1. Introduction

MIFTI is a device being planned by US private company Magneto-Inertial Fusion

Technologies, Inc [61].

5.29.2. Purpose

The purpose of MIFTI is to research and develop the pinch line for fusion energy production.

5.29.3. Main features

MIFTI is being designed as a pinch device. General information is listed in Table 120 below.

TABLE 120. GENERAL INFORMATION

Device type

Pinch

Status

Planned

Ownership

Private

136

5.30. PFRC (PRINCETON FUSION SYSTEMS, UNITED STATES OF AMERICA)

5.30.1. Introduction

PFRC is a device being planned by US private company Princeton Satellite Systems [61].

5.30.2. Purpose

The purpose of PFRC is to demonstrate the feasibility of the field reversed configuration as a

portable and modular fusion power plant.

5.30.3. Main features

PFRC is a fusion microreactor concept based on the field reversed configuration. It is expected

to be portable (sized to fit on a truck) and produce power ranging 1–10 MW. The concept is

expected to be applicable for space applications as a thrust-controlled rocket, or as direct fusion

drive. General information is listed in Table 121 below.

TABLE 121. GENERAL INFORMATION

Device type

Field Reversed Configuration

Status

Planned

Ownership

Private

137

5.31. Z MACHINE (SANDIA NATIONAL LABORATORIES, UNITED STATES OF

AMERICA)

5.31.1. Introduction

Z machine is located at Sandia National Laboratories, USA. It is a part of a research program

launched in 1960. Z machine is the most powerful laboratory radiation source in the world.

5.31.2. Purpose

The purpose of Z machine is to support the development of magnetized liner inertial fusion.

5.31.3. Main features

Z machine is a pinch fusion device. It uses capacitors to activate powerful electrical pulses.

These hit a small target made of hundreds tungsten wires placed in a hohlraum (a small metal

container) located in the centre of the machine. The deposited energy creates a strong magnetic

field that pushes the exploded particles inside to collide. The hohlraum walls get heated up to

1.8 million degrees Celsius by the radiation resulting from collisional processes [86]. General

information is listed in Table 122 below.

TABLE 122. GENERAL INFORMATION

Device type

Pinch

Status

Operating

Ownership

Public

138

5.32. NORMAN (C-2W) (TAE TECHNOLOGIES, UNITED STATES OF AMERICA)

5.32.1. Introduction

Operating since 2017, Norman is the fifth device developed and operated by US private

company TAE Technologies [61].

5.32.2. Purpose

The purpose of Norman is to show plasma ramp-up by neutral beam injection and current drive

in a reversed field configuration, as well as to improve edge and divertor plasma performance,

obtaining plasma temperature up to 3 keV.

5.32.3. Main features

Norman is a field reversed configuration device. Plasma heating and current drive is achieved

via neutral beam injection. Norman is the largest theta-pinch collisional-merging system in the

world. It consists of a central confinement section surrounded by 2 internal divertors; 2 sections

for field reversed theta-pinch formation; and 2 outer divertors. Liquid nitrogen is used for the

cooling system, improving pumping performance inside the divertors [87]. General information

is listed in Table 123 below.

TABLE 123. GENERAL INFORMATION

Device type

Field Reversed Configuration

Status

Operating

Ownership

Private

139

5.33. COPERNICUS (TAE TECHNOLOGIES, UNITED STATES OF AMERICA)

5.33.1. Introduction

Copernicus is the next step device being constructed by US private company TAE Technologies

[61].

5.33.2. Purpose

The purpose of Copernicus is to demonstrate net scientific energy gain

in a reversed field

configuration by 2025, simulating the D-T fuel cycle performance while using only hydrogen

fuel.

5.33.3. Main features

Copernicus will be a field reversed configuration device operating with hydrogen plasma. It is

expected to become operational by 2024. General information is listed in Table 124 below.

TABLE 124. GENERAL INFORMATION

Device type

Field Reversed Configuration

Status

Under construction

Ownership

Private

140

5.34. ZEBRA (UNIVERSITY OF NEVADA, UNITED STATES OF AMERICA)

5.34.1. Introduction

Operating since 2000, Zebra is located at University of Nevada, USA.

5.34.2. Purpose

The purpose of Zebra is to carry out research as well as contribute to training students in the

field of high energy density science.

5.34.3. Main features

Zebra is a pulsed power generator [88]. General information is listed in Table 125 below.

TABLE 125. GENERAL INFORMATION

Device type

Pinch

Status

Operating

Ownership

Public

141

5.35. MST (UNIVERSITY OF WISCONSIN-MADISON, UNITED STATES OF

AMERICA)

5.35.1. Introduction

MST is located at University of Wisconsin-Madison, USA.

5.35.2. Purpose

The main purpose of MST is to advance plasma physics research.

5.35.3. Main features

MST is a reversed field pinch, which can also operate with tokamak geometries. MST features

an aluminium shell that, depending on the chosen configuration, can serve either as vacuum

vessel or as equilibrium magnet [89]. General information is listed in Table 126 below.

TABLE 126. GENERAL INFORMATION

Device type

Reversed Field Pinch

Status

Operating

Ownership

Private

142

5.36. FUZE-Q (ZAP ENERGY INC., UNITED STATES OF AMERICA)

5.36.1. Introduction

Operating since 2018, FuZE-Q is a device designed and operated by US private company Zap

Energy Inc. [61].

5.36.2. Purpose

The purpose of FuZE-Q is to reach scientific energy gain

sci

=1.

5.36.3. Main features

FuZE-Q is a pinch. An electric current generates sheared flows. These produce a magnetic field

that confines and compresses the plasma. The Z-pinch plasma is heated and compressed by

flowing an extremely large current (~10

amps) through the plasma. General information is

listed in Table 127 below.

TABLE 127. GENERAL INFORMATION

Device type

Pinch

Status

Operating

Ownership

Private

143

6. DEMO DEVICES

6.1. CFETR (CHINESE CONSORTIUM, CHINA)

6.1.1. Introduction

CFETR is a DEMO concept based on conventional tokamak design being developed in China

by a Chinese Consortium. CFETR is the next device in the roadmap for the realization of fusion

energy in China. The conceptual design was completed in 2015. Construction of the CFETR is

expected to be completed by 2040.

6.1.2. Purpose

CFETR is expected to bridge the gaps between ITER and a fusion power plant, as well as to

demonstrate net engineering gain

eng

>1).

6.1.3. Main features

CFETR R&D plan is expected to consist of two phases. During the first phase, the efforts will

focus on achieving steady-state operation and tritium self-sufficiency with fusion power up to

200 MW. This phase will address issues relevant to burning plasma physics R&D in order to

demonstrate steady-state advanced operation. The second phase will focus on validating

DEMO-relevant issues with fusion power above 1 GW [90]. Some of CFETR conceptual

features are: major radius of 7.2 m, minor radius of 2.2 m, plasma elongation κ

=2, plasma

current of 14 MA, magnetic field on axis of 6.5 T, normalized beta β

=2.3, a predicted scientific

energy gain

sci

=30. General information is listed in Table 128 below.

TABLE 128. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Public

144

6.2. EU-DEMO (EUROFUSION, EUROPEAN UNION)

6.2.1. Introduction

EU-DEMO is a DEMO concept based on conventional tokamak design being developed in

Europe by EUROfusion. The European DEMO or EU-DEMO is the facility expected to follow

ITER in the European roadmap to electricity from fusion. The project is currently in its

conceptual design phase and is expected to start operating by 2050.

6.2.2. Purpose

EU-DEMO aims to demonstrate the technological and economic viability of fusion by

producing about 500 MW of net electricity and to achieve tritium self-sufficiency.

6.2.3. Main features

Several design options are being studied. These options will have an impact on a number of

plant technologies, including the divertor configuration and breeding blanket solution, among

others. The pre-conceptual design of EU-DEMO tokamak foresees a major radius of ∼9 m and

a fusion power of ∼2000 MW [91]. General information is listed in Table 129 below.

TABLE 129. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Public

145

6.3. JA-DEMO (JAPANESE CONSORTIUM, JAPAN)

6.3.1. Introduction

The Japanese DEMO or JA-DEMO is a DEMO concept based on conventional tokamak design

being developed in Japan by a Japanese Consortium. Construction of JA-DEMO is expected to

be completed by 2050.

6.3.2. Purpose

The Purpose of JA-DEMO are to demonstrate net engineering gain

eng

>1) and tritium self-

sufficiency, as well as plant availability bridging the gap to commercialization of fusion energy.

6.3.3. Main features

It is expected that for reliable electric power generation, a fusion output of 1.5 GW or higher,

will be required. For the magnets system of JA-DEMO, superconducting (Nb3Sn) magnets

consisting of a central solenoid, 7 poloidal field coils (NbTi), 16 toroidal field coils (Nb3Sn or

Nb3Al) are being considered. Some of JA-DEMO conceptual features are: major radius of 8.5

m, minor radius of 2.42 m, plasma elongation κ

=1.65, plasma current of 12.3 MA, magnetic

field on axis of 5.94 T, normalized beta β

=3.4, current drive power of 83.7 MW and a predicted

scientific energy gain

sci

=17.5 [92]. General information is listed in Table 130 below.

TABLE 130. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Public

146

6.4. K-DEMO (KOREA INSTITUTE OF FUSION ENERGY, REPUBLIC OF KOREA)

6.4.1. Introduction

K-DEMO is a DEMO concept based on conventional tokamak design being developed in the

Republic of Korea by the Korean Institute of Fusion Energy. Construction of K-DEMO is

expected to be fully completed by 2050. In a first early phase of the project (2037–2050), K-

DEMO will be used to develop and test components. In its second phase, after 2050, it is

expected to demonstrate net electrical power.

6.4.2. Purpose

The Purpose of K-DEMO are to demonstrate physics and technology necessary for achieving

net engineering gain

eng

>1).

6.4.3. Main features

The conceptual design of K-DEMO features a major radius of 6.8 m, minor radius of 2.1 m,

toroidal field of about 7 T, and plasma current larger than 12 MA. K-DEMO is expected to

feature a double-null divertor configuration and the divertor X-point inside the vacuum vessel.

The K-DEMO blanket sectors are subdivided into 16 inboard and 32 outboard sectors. The

upper or lower divertor is also subdivided into 32 modules. The key features of the K-DEMO

magnet system include two toroidal field coil winding packs with different conductors, enclosed

in the toroidal field case [93]. General information is listed in Table 131 below.

TABLE 131. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Public

147

6.5. DEMO-RF (RUSSIAN CONSORTIUM, RUSSIAN FEDERATION)

6.5.1. Introduction

DEMO-RF is a DEMO concept based on conventional tokamak design being developed in the

Russian Federation by a Russian Consortium. Construction of DEMO-RF is expected to be

completed by 2055.

6.5.2. Purpose

DEMO-RF is expected to demonstrate net engineering gain

eng

>1).

6.5.3. Main features

DEMO-RF features are under development. The conceptual design currently foresees the use

of the facility either as a pure fusion energy system or as a fusion–fission hybrid facility with

high temperature superconducting magnets and a total magnetic field larger than 8 T and plasma

current of about 5 MA. Liquid metal plasma facing components are being considered for first

wall and divertor [94]. General information is listed in Table 132 below.

TABLE 132. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Public

148

6.6. FDP (GENERAL FUSION INC, UNITED KINGDOM)

6.6.1. Introduction

FDP is a DEMO concept based on magnetized target fusion design (which uses pneumatic

pistons to compress the plasma) being developed by Canadian company General Fusion

(supported by the government of Canada) [61]. Its target completion date is 2025.

6.6.2. Purpose

The purpose of the FDP is to prove that magnetized target fusion technology can scale to a

commercial pilot plant.

6.6.3. Main features

The three main components of FDP are a liquid metal chamber, compression system and plasma

injector. The device is expected to operate by heating hydrogen atoms at high temperatures and

compressing it with pneumatic pistons surrounding a rotating internal chamber filled with liquid

metal. The generated heat from fusion reactions can then be transferred into the liquid metal,

and in future commercial power plants, it can be extracted from the metal and used to produce

steam, which will drive a turbine producing electricity. General information is listed in Table

133 below.

TABLE 133. GENERAL INFORMATION

Device type

Magnetized Target Fusion

Status

Planned

Ownership

Public-Private

149

6.7. ST-E1 (TOKAMAK ENERGY, UNITED KINGDOM)

6.7.1. Introduction

ST-E1 is a DEMO concept based on spherical tokamak design being developed by UK company

Tokamak Energy Ltd [61]. ST-E1 is expected to be completed by 2030.

6.7.2. Purpose

The purpose of ST-E1 is to demonstrate net engineering gain

eng

>1).

6.7.3. Main features

ST-E1 will be a compact spherical tokamak with high temperature superconducting magnets.

General information is listed in Table 134 below.

TABLE 134. GENERAL INFORMATION

Device type

Spherical Tokamak

Status

Planned

Ownership

Private

150

6.8. STEP (UKAEA, UNITED KINGDOM)

6.8.1. Introduction

STEP is a DEMO concept based on spherical tokamak design being developed by UKAEA.

The first phase of the programme is to produce a concept design by 2024. Its target completion

date is 2040. STEP is expected to be smaller than ITER. The spherical shape can improve

efficiency in the magnetic field and potentially minimise the plant’s costs.

6.8.2. Purpose

The STEP programme aims to demonstrate the commercial viability of fusion, as well as to

enable the flourishing of a fusion industry.

6.8.3. Main features

STEP will be a compact spherical tokamak able to produce net engineering gain

eng

>1),

although it is not expected to be a commercially operating plant [95]. General information is

listed in Table 135 below.

TABLE 135. GENERAL INFORMATION

Device type

Spherical Tokamak

Status

Planned

Ownership

Public

151

6.9. ARC (COMMONWEALTH FUSION SYSTEMS, UNITED STATES OF AMERICA)

6.9.1. Introduction

ARC is a DEMO concept based on conventional tokamak design being developed by US private

company Commonwealth Fusion Systems [61]. ARC will be the successor of SPARC (see p.

77).

6.9.2. Purpose

The purpose of ARC is to demonstrate the commercial viability of fusion with high temperature

superconducting magnet technology.

6.9.3. Main features

ARC will be a compact conventional tokamak with high temperature superconducting magnets

able to produce ∼ 200–250 MWe, with a radius of 3.3 m, a minor radius of 1.1 m, and an on-

axis magnetic field of 9.2 T. ARC will feature REBCO superconducting toroidal field coils.

The coils will have joints to enable disassembly, which will allow for quick replacements of

the vacuum vessel (thus mitigating first wall life-time issues) and enable the possibility of

testing various vacuum vessel designs and divertor materials [96]. General information is listed

in Table 138 below.

TABLE 138. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Private

152

6.10. GA-FPP (GENERAL ATOMICS, UNITED STATES OF AMERICA)

6.10.1. Introduction

The GA-FPP is a DEMO concept based on steady-state, compact advanced tokamak design,

which was announced in October 2022 [97].

6.10.2. Purpose

The purpose of the GA-FPP is to demonstrate the commercial viability of fusion, achieving

steady-state operation, maximizing efficiency, reducing maintenance costs and increasing the

lifetime of the facility.

6.10.3. Main features

The GA-FPP design approach will rely on advanced sensors, control algorithms and high

performance computers for controlling the plasma, silicon-carbide breeding blankets for

producing the necessary Tritium and microwave heating for powering the fusion reactions.

TABLE 139. GENERAL INFORMATION

Device type

Conventional Tokamak

Status

Planned

Ownership

Private

153

6.11. DA VINCI (TAE TECHNOLOGIES, UNITED STATES OF AMERICA)

6.11.1. Introduction

Da Vinci is a DEMO concept based on field reversed configuration design being developed by

US private company TAE Technologies [61]. Da Vinci will be the successor of Copernicus (see

p. 139).

6.11.2. Purpose

The purpose of Da Vinci is to demonstrate the commercial viability of fusion in a reversed field

configuration via p-

B reactions (see Table 1, p. 4).

6.11.3. Main features

Da Vinci will be a field reversed configuration device. General information is listed in Table

140 below.

TABLE 140. GENERAL INFORMATION

Device type

Field Reversed Configuration

Status

Planned

Ownership

Private

155

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161

CONTRIBUTORS TO DRAFTING AND REVIEW

M. Barbarino

International Atomic Energy Agency

B. Bigot

ITER Organization

I. Chapman

United Kingdom Atomic Energy Authority, UK

S. Chaturvedi

Institute for Plasma Research, India

A.J.H. Donné

EUROfusion

L.-G., Eriksson

European Commission

S. Günter

Max Planck Institute for Plasma Physics, Germany

M. Hole

Australian National University, Australia

J.G. Jacquinot

ITER Organization

B.V. Kuteev

National Research Centre Kurchatov Institute, Russian

Federation

Y. Liu

Southwestern Institute of Physics, China

H. Park

Ulsan National Institute of Science & Technology, Republic of

Korea

D. Ridikas

International Atomic Energy Agency

J. Sanchez Sanz

CIEMAT, Spain

I. Tazhibayeva

National Nuclear Center, Kazakhstan

E. Tsitrone

French Alternative Energies and Atomic Energy Commission,

France

Y. Ueda

Osaka University, Japan

J. Van Dam

United States Department of Energy, USA

I. Vargas-Blanco

Instituto Tecnologico de Costa Rica, Costa Rica

N. Yalynskaya

International Atomic Energy Agency

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