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JT-60
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JT-60
Japan Torus-60
Device typeTokamak
LocationNaka, Ibaraki Prefecture, Japan
AffiliationJapan Atomic Energy Agency
Technical specifications
Major radius3.4 m (11 ft)
Minor radius1.0 m (3 ft 3 in)
Plasma volume90 m3
Magnetic field4 T (40,000 G) (toroidal)
Discharge duration65 s
History
Year(s) of operation1985–2010
Preceded byJFT-2M
Succeeded byJT-60SA
Related devicesTFTR
Links
Websitewww.qst.go.jp/site/jt60-english/
JT-60SA
Japan Torus-60 Super Advanced
Device typeTokamak
LocationNaka, Ibaraki Prefecture, Japan
AffiliationQST + F4E
Technical specifications
Discharge duration100 s
History
Date(s) of construction2013–2020
Year(s) of operation2023–present
Preceded byJT-60U
Related devicesITER
Links
Websitewww.jt60sa.org/wp/

JT-60 (short for Japan Torus-60) is a large research tokamak, the flagship of the Japanese National Institute for Quantum Science and Technology's fusion energy directorate. As of 2023 the device is known as JT-60SA and is the largest operational superconducting tokamak in the world,[1] built and operated jointly by the European Union and Japan in Naka, Ibaraki Prefecture.[2][3] SA stands for super advanced tokamak, including a D-shaped plasma cross-section, superconducting coils, and active feedback control.

JT-60 claimed that it held the record[a] for the highest value of the fusion triple product achieved: 1.77×1028 K·s·m−3 = 1.53×1021 keV·s·m−3.[4][5] The product quoted is not a valid fusion triple product since the plasmas did not satisfy the steady state of the Lawson criterion as discussed below.

JT-60 also claimed without proof that it held the record[a] for the hottest ion temperature ever achieved (522 megakelvins). In reality the TFTR machine at Princeton routinely measured higher ion temperatures during the 1993-1996 campaign, as discussed below.[6]

Original design

[edit]

JT-60 was first designed in the 1970s during a period of increased interest in nuclear fusion from major world powers. In particular, the US, UK and Japan were motivated by the excellent performance of the Soviet T-3 in 1968 to further advance the field. The Japanese Atomic Energy Research Institute (JAERI), previously dedicated to fission research since 1956, allocated efforts to fusion.

JT-60 began operations on April 8, 1985,[7] and demonstrated performance far below predictions, much like the TFTR and JET that had begun operations shortly prior.

Over the next two decades, TFTR, JET and JT-60 led the effort to regain the performance originally expected of these machines. JT-60 underwent a major modification during this time, JT-60U (for "upgrade") in March 1991.[8] The change resulted in significant improvements in plasma performance.

JT-60/TFTR disputed records

[edit]

By 1996, JT-60 had achieved its record ion temperature of 45 keV,[6] which is claimed to have exceeded the highest temperatures measured at that time in the TFTR tokamak in Princeton. Detailed measurements of the ion temperatures analyzed during TFTR's experimental campaign with deuterium-tritium plasmas in 1993–1996, found numerious discharges with temperatures greater than 50 keV in both deuterium-only and deuterium-tritium plasmas.[9] A 2025 publication of a reanalysis of TFTR transport and confinement results for a selected scan of discharges mentions that several "supershots", not in the scan, had ion temperatures of 70 keV with a measurement error bar of 28%.[9]

The TFTR team did not highlight these high temperatures for several reasons. The ion temperature measurements in JT-60, TFTR, and JET measured only singly ionized trace carbon impurity ions, not the temperatures of the hydrogenic ions. The carbon ions do not fuse, and displace the deuterium and tritium ions which can fuse. The hydrogenic ion temperatures can be calculated in the TRANSP analysis code. The methods used are published and widely used in analysis of experimental results. [10] These temperatures are the relevant ones for calculating the deuterium and tritium fusion reactions. They generally are less than the carbon temperatures. Secondly, the end goal of this research, practical minimally poluting fusion energy, does not require ion temperatures greater than about 25 keV. An example of simulation of a burning plasma in ITER is [11]

The fusion triple product metric applies only to plasmas in steady state, as stated explicitly in the Lawson criterion. The JT-60 plasmas with high values were far from steady state; in fact, their conditions rose rapidly in time to those values, and then suffered major disruptions, which extinguished the plasmas abruptly. Examples are in. [12] [13] Also the derivation of the fusion triple product assumes that the fusion power results from thermonuclear fusion (from thermal deuterium and tritium). Instead the high fusion power in past tokamak experiments resulted dominatly from beam-thermal reactions.

Thus the JT-60's claimed record for the triple product is not a 'fusion triple product'. Steady state discharges have been achieved in other devices such as Tore Supra and WEST have announced results for the fusion triple product.[14]

JT-60U (Upgrade)

[edit]

The main objective of the JT-60U upgrade was to "investigate energy confinement near the breakeven condition, [a] non-inductive current drive and burning plasma physics with deuterium plasmas." To accomplish this, the poloidal field coils and the vacuum vessel were replaced. Construction began in November 1989 and was completed in March 1991.[15] Operations began in July.[16]

JT-60U researchers claimed that on October 31, 1996, they achieved an estimated breakeven factor of QDTeq = 1.05 at 2.8 MA.[17] In other words, if the homogenous deuterium fuel was theoretically replaced with a 1:1 mix of deuterium and tritium, the fusion reaction is estimated to have created an energy output 1.05 times the energy injected into the tokamak. An estimate based on a discharge in 1968 gave QDTeq = 1.25.[18] The record of the central ratio Qcore achieved in a tokamak discharge is 1.3 in JET in 1998. [19]

A credible estimate of extrapolation of a deuterium plasma to a deuterium-tritium plasma requires starting with a validated and verified integrated computer model, and then reruning with a deuterium-tritium mixture to calculate the fusion yield. Details of the deuterium plasma also need to be shown for credibility. An example of such an estimate was published before TFTR started its deuterium-tritium campaign in 1993–1996.[20] This paper calculated that the QDTeq would be 0.32. In retrospect, the record achieved was 0.28 (discharge 80539) so the projection were optimistic. A much larger amount of energy was injected into the TFTR and JT-60U test chambers. JT-60U was not equipped to utilize tritium, as it would add extensive costs and safety risks.[b]

In February 1997, a modification to the divertor from an open-type shape to a semi-closed W-shape for greater particle and impurity control was started and later completed in May.[21][22][23] Experiments simulating the helium exhaust in ITER were promptly performed with the modified divertor, with great success. In 1998, the modification allowed JT-60U to reach an estimated fusion energy gain factor of QDTeq = 1.25 at 2.6 MA,[24][25][26] as discussed above.

In December 1998, a modification to the vacuum pumping system that began in 1994 was completed. In particular, twelve turbomolecular pumps with oil bearings and four oil sealed rotary vacuum pumps were replaced with magnetically suspended turbomolecular pumps and dry vacuum pumps. The modification reduced the 15-year-old system's consumption of liquid nitrogen by two thirds.[27]

In fiscal year 2003, the plasma discharge duration of JT-60U was successfully extended from 15 s to 65 s.[28]

In 2005, ferritic steel (ferromagnet) tiles were installed in the vacuum vessel to correct the magnetic field structure and hence reduce the loss of fast ions.[29][30] The JAEA used new parts in the JT-60, having improved its capability to hold the plasma in its powerful toroidal magnetic field.

Sometime in 2007-2008, in order to control plasma pressure at the pedestal region and to evaluate the effect of fuel on the self-organization structure of plasma, a supersonic molecular beam injection (SMBI) system was installed in JT-60U. The system's design was a collaboration between Cadarache, CEA, and JAEA.[31] QDTeq JT-60U ended operations on August 29, 2008.[32]

JT-60SA

[edit]
JT-60SA under construction in 2016.

JT-60SA is the successor to JT-60U, operating as a satellite to ITER as described by the Broader Approach Agreement. It is a fully superconducting tokamak with flexible components that can be adjusted to find optimized plasma configurations and address key physics issues.[33] Assembly began in January 2013 and was completed in March 2020. After a major short circuit during integrated commissioning in March 2021 necessitating lengthy repairs, it was declared active on December 1, 2023. The overall cost of its construction is estimated to be around €560000000, adjusted for inflation.[34]

Weighing roughly 2,600 short tons (2,400 t),[35] JT-60SA's superconducting magnet system includes 18 D-shaped niobium-titanium toroidal field coils, a niobium-tin central solenoid, and 12 equilibrium field coils.

History

[edit]

The idea of an advanced tokamak, a tokamak utilizing superconducting coils, traces back to the early 1960's. The idea seemed very promising, but was not without its problems. Around January 1972, engineers at JAERI initiated an effort to further research the idea and try to solve its hurdles.[36] This initiative progressed in parallel with the development of JT-60,[37] and by 1983-84 it was decided that it constituted its own experimental reactor: FER (Fusion Experimental Reactor).[38]

However, the JT-60U upgrade in 1991 demonstrated the significant flexibility of the JT-60 facilities and assembly site, so by January 1993 FER was designated as a modification to JT-60U and renamed JT-60SU (for Super Upgrade).[39]

In January 1996, a paper detailing the superconducting properties of Nb3Al composite wire and its fabrication process was published in the 16th International Cryogenic Engineering/Materials Conference journal.[40] Engineers assessed the potential use of the aluminide in JT-60SU's 18 toroidal coils.[41]

Designs and intentions for the modification varied over the next decade, until February 2007, when the Broader Approach Agreement was signed between Japan and the European Atomic Energy Community.[42] In it, the Satellite Tokamak Program established a clear, defined goal for JT-60SA: act as a small-scale ITER. This way, JT-60SA could give hindsight to engineers assembling and operating the full-scale reactor in the future.

It was planned for JT-60 to be disassembled and then upgraded to JT-60SA by adding niobium-titanium superconducting coils by 2010.[4][43] It was intended for the JT60SA to be able to run with the same shape plasma as ITER.[43]: 3.1.3  The central solenoid was designed to use niobium-tin (because of the higher (9 T) field).[43]: 3.3.1 

Assembly

[edit]

Construction of the tokamak officially began on 28 January 2013 with the assembly of the cryostat base, which was shipped from Avilés, Spain over a 75-day-long journey.[c] The event was highly publicized through local and national news, and reporters from 10 media organizations were able to witness it in person.[44]

Assembly of the vacuum vessel began in May 2014. The vacuum vessel was manufactured as ten sectors with varying arcs (20°×1, 30°×2, 40°×7) that had to be installed sequentially. On June 4, 2014, two of ten sectors were installed. In November 2014 seven sectors had been installed. In January 2015 nine sectors had been installed.

Construction was to continue until 2020 with first plasma planned in September 2020.[45] Assembly was completed on March 30, 2020,[46] and in March 2021 it reached its full design toroidal field successfully, with a current of 25.7 kA.[47]

Short circuit

[edit]

On March 9, 2021, a coil energization test was being performed on equilibrium field coil no. 1 (EF1) when the coil current rapidly increased, then suddenly flatlined. The reactor was safely shut down over the next few minutes, during which the pressure in the cryostat increased from 10×10−3 Pa to 7000 Pa. Investigations immediately followed.

The incident, which came to be known as the "EF1 feeder incident", was found to be caused by a major short circuit resulting from insufficient insulation of the quench detection wire conductor exit. The formed arc damaged the shells of EF1, causing a helium leak to the cryostat.

In total, 90 locations required repairs and machine sensors needed to be rewired. However, the intricate JT-60SA was designed and assembled with intense precision, meaning access to the machine was sometimes limited. Risks of further delay to plasma operations compounded the issue.[48]

The JT-60SA team was disappointed with the incident, given how close the machine was to operation, but persevered.

Repairs were completed in May 2023 and preparations for operation began.[49]

Present operations

[edit]

JT-60SA achieved first plasma on October 23, 2023, making it the largest operational superconducting tokamak in the world as of 2024.[1] The reactor was declared active on December 1, 2023.[50]

Specifications

[edit]

(60 stands for JT-60, 60U stands for JT-60U, 60SA stands for JT-60SA) ("60SA I" refers to the initial/integrated research phase of JT-60SA, "60SA II" refers to the extended research phase)

Plasma[51][52][53]
Volume Current Major radius Minor radius Aspect ratio Height Pulse length Elongation Triangularity
60 2.1 MA - 2.6 MA 3 m 0.85 m - 0.95 m 3.52 - 3.15 5 s
60U 90 m3 3 MA 3.4 m 1 m 3.4 1.5±0.3 m 65 s 1.5±0.3
60SA I 5.5 MA 2.97 m 1.17 m 2.54 2.14 m 100 s 1.83 0.50
60SA II 5.5 MA 2.97 m 1.18 m 2.52 2.28 m 100 s 1.93 0.57
Vacuum Vessel[54][52]
Material Baking temp. One-turn resistance
60 Inconel 625 500 °C > 1.3 mΩ
60U Inconel 625 300 °C 0.2 mΩ
60SA SS 316L 200 °C 16 µΩ
Toroidal Field Coils[52]
# Turns Material Coil current Inductance Resistance Time constant
60 18 1296 52.1 kA 2.1 H 84 mΩ 25 s
60U 18 1296 AgOFCu 52.1 kA 2.1 H 97 mΩ 21.65 s
60SA

Notes

[edit]

References

[edit]
[edit]
Revisions and contributorsEdit on WikipediaRead on Wikipedia

JT-60

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JT-60 is a prominent tokamak-type research device developed and operated by the Japan Atomic Energy Research Institute (JAERI, now part of the Japan Atomic Energy Agency or JAEA) at the Naka Fusion Institute in , . Constructed in response to global energy challenges in the 1970s, it began producing its first plasma in April 1985 and conducted experiments until its decommissioning in August 2008, spanning over two decades of operation. Designed to investigate plasma confinement, heating, and current drive in reactor-relevant hydrogen plasmas, JT-60 featured a D-shaped vacuum vessel, neutral beam injection for heating up to tens of megawatts, and capabilities for long-pulse operations of 5–10 seconds, enabling studies toward fusion break-even conditions (equivalent Q ≥ 1). Throughout its operational life, JT-60 achieved several groundbreaking milestones in fusion science, including the world's highest central of 45 keV in 1996—recognized in the Guinness Book of Records—and the highest central of 20 keV at the time. It set a record for the fusion (a key metric combining plasma , , and confinement time) of 1.53 × 10^{21} keV·s·m^{-3}, surpassing other devices and providing critical data for the International Thermonuclear Experimental Reactor () design. The device also pioneered advanced regimes, such as reversed shear plasmas and high-bootstrap current operations, which demonstrated potential for steady-state production with reduced external current drive. These results advanced understanding of magnetohydrodynamic stability, impurity transport, and energetic particle behavior in high-performance plasmas. In 1991–1992, JT-60 underwent a major upgrade to JT-60U, enhancing its divertor configuration, increasing plasma shaping flexibility (elongation up to 1.8), and boosting heating power to over 40 MW via neutral beam and radiofrequency systems, allowing plasma currents up to 5 MA and toroidal fields up to 4.2 T. This iteration further contributed to by validating high-beta operations and long-pulse scenarios, with equivalent energy confinement times exceeding those projected for 's baseline. Following its closure, the JT-60 facility was transformed into JT-60SA, a joint Japan-Europe superconducting that achieved first plasma in October 2023, reached 1 MA plasma current in December 2023, and was certified by as holding the largest plasma volume in September 2024. As of 2025, it is undergoing upgrades for operations resuming in 2026, building directly on JT-60's legacy to support DEMO reactor development and commercial fusion pathways.

Background

Purpose in fusion research

JT-60 is a prominent device designed to advance research by confining high-temperature plasmas using toroidal and poloidal magnetic fields generated by external coils, enabling the study of fusion reactions in a doughnut-shaped vacuum vessel. The original JT-60 aimed to achieve energy conditions (equivalent Q ≥ 1) and establish the scientific basis for fusion reactors through investigations of plasma confinement, heating, and current drive in reactor-relevant plasmas. As one of the world's largest tokamaks, it followed major devices such as the (JET) in and the Tokamak Fusion Test Reactor (TFTR) in the United States, which in the 1980s established key milestones in plasma heating and confinement. JT-60's development positioned it as a critical platform for international collaboration, contributing to the scientific foundation for subsequent projects like by demonstrating scalable plasma behaviors in larger volumes. The primary goals of the original JT-60 focused on exploring plasma regimes toward , including studies of high-performance confinement and magnetohydrodynamic (MHD) stability. Over its evolution, JT-60 transitioned from a copper-coil operational in the mid-1980s to the upgraded JT-60U in the early 1990s, which enhanced divertor capabilities and heating power, and finally to the fully superconducting JT-60SA, achieving first plasma in 2023 as part of the Japan-Europe Broader Approach agreement. This progression reflects a strategic shift toward long-pulse, high-field operations mimicking ITER's design, with JT-60SA serving as its key satellite for pre-ITER validation. JT-60SA's research plan encompasses the exploration of high-performance plasma regimes, non-inductive current drive techniques, and MHD stability to validate operational scenarios for and future fusion reactors, including advanced modes with internal transport barriers and high bootstrap currents to achieve sustained fusion conditions, while addressing challenges like edge-localized modes (ELMs) and disruption mitigation through optimized current profiles and heating systems. These efforts support by providing complementary data on steady-state operations and high-beta plasmas, essential for efficient energy production without continuous external power input. As of 2025, JT-60SA represents the largest operational superconducting globally, uniquely equipped to support DEMO reactor concepts through studies of integrated power exhaust, metallic wall compatibility, and full non-inductive scenarios up to 100 seconds. Its capabilities, including advanced neutral beam and electron cyclotron systems, enable pioneering research on economically viable fusion paths beyond .

Historical development

The development of JT-60 originated in the amid 's national push for fusion energy research, spurred by global oil crises and the need for alternative energy sources. The Atomic Energy Research Institute (JAERI), established as the primary institution for atomic energy studies, initiated the project as a large-scale to pursue breakeven plasma conditions. commenced in the late at the Naka site, with manufacturing activities ramping up by under JAERI's oversight. This effort was funded through 's Science and Technology Agency as part of the second phase of its fusion research and development program. The original JT-60 achieved its first plasma on April 8, 1985, marking a significant in Japan's fusion endeavors and establishing it as one of the world's largest at the time. Internationally, JT-60 participated in the (IEA) Implementing Agreement on Cooperation Among Large Facilities, signed in 1986, which fostered collaboration with Europe's JET and the US's TFTR to advance research efficiency and data sharing. By 1991, operations of the original JT-60 concluded to facilitate its upgrade to JT-60U, which began the same year and extended the facility's capabilities for higher plasma . Institutionally, JT-60's management evolved with broader reorganizations in 's fusion sector. JAERI merged into the Japan Atomic Energy Agency (JAEA) in , placing the Naka facility under JAEA's Naka Fusion Institute. In 2016, the institute transitioned to the National Institutes for Quantum Science and Technology (QST), continuing oversight of JT-60 operations and upgrades. A pivotal shift occurred in 2007 with the Broader Approach Agreement, signed on February 5 between and (effective June 1), which integrated JT-60 into -EU fusion collaboration to support and future DEMO reactors; this agreement designated the upgrade to JT-60SA as a key project hosted at the Naka site. Key subsequent milestones included the end of JT-60U operations in August 2008, followed by decommissioning of its copper coils around 2010 as part of the transition to superconducting components. Assembly of JT-60SA commenced on January 28, 2013, with the installation of the cryostat base, progressing through international and integration efforts thereafter. These developments underscored JT-60's enduring role in global fusion progress under evolving institutional and collaborative frameworks.

Original JT-60

Design and construction

The construction of the original JT-60 tokamak commenced in 1977 at the Naka Fusion Research Establishment, operated by the Japan Atomic Energy Research Institute (now part of the National Institutes for Quantum Science and Technology), and was completed in 1985, with a total cost of approximately 230 billion yen. This multi-year effort involved collaboration with Japanese industry for the development and fabrication of major components, marking a significant investment in Japan's fusion research program during the late 1970s and early 1980s. The design emphasized a large-scale, non-superconducting configuration optimized for high-performance plasma experiments, featuring a large-bore vacuum vessel with a major of 3.04 m to accommodate extended plasma volumes and a circular plasma cross-section with an outboard equatorial divertor. Central to the inductive current drive system was an air-core , which provided the necessary for plasma initiation and sustainment without the limitations of iron-core saturation, enabling pulse lengths up to several seconds. The poloidal field (PF) coils, constructed from and water-cooled, were arranged to shape and position the plasma, supporting versatile operational modes including single-null divertor configurations. The vacuum vessel was fabricated from to ensure structural integrity under high vacuum and thermal loads, incorporating a limiter to manage and protect vessel walls during plasma contact. Key components, including the limiter and PF coils, featured water-cooling systems to handle the substantial power dissipation during operations. Engineering challenges centered on attaining a high (approximately 2.7) while maximizing flux swing from the air-core to facilitate long-pulse discharges without excessive resistive losses.

Initial operations and achievements

The JT-60 achieved its first plasma on April 8, 1985, initiating a series of ohmic heating experiments that explored basic plasma behavior and confinement in the large-scale device. Initial operations focused on establishing stable discharges with plasma currents up to 1.6 MA in both and divertor configurations, providing foundational data on energy confinement and impurity under plasma conditions. These early runs, conducted from April to June 1985, validated the machine's engineering design, including its outboard divertor and poloidal field coils, while achieving line-averaged electron densities around 5 × 10¹⁹ m⁻³. By 1988, neutral beam injection enabled access to the high-confinement H-mode regime, a key advancement that improved energy confinement times by forming an edge transport barrier. Notable achievements included plasma currents reaching 3.2 MA in discharges with neutral beam heating powers up to 20 MW, alongside a record normalized beta value of approximately 4.8% in high-performance shots, highlighting the device's capability for reactor-relevant pressure conditions. Additionally, long-pulse edge-localized mode ()-suppressed H-mode discharges were sustained for up to 12 seconds, demonstrating initial progress toward extended operation with controlled edge stability. Experimental efforts emphasized divertor performance for heat and particle exhaust, with studies revealing effective impurity screening in lower X-point configurations introduced in 1988. Pellet injection experiments provided precise control, enabling peaked profiles that enhanced confinement and allowed exploration of limit scaling in high-current plasmas. Validation of the neoclassical bootstrap current was also pursued through magnetic measurements in high-beta discharges, confirming up to 20-30% of the total current arising from pressure-driven effects as predicted by theory. Despite these successes, operations from 1985 to 1991 were constrained to inductive current drive, limiting pulse durations to seconds and preventing steady-state studies essential for reactor design, which ultimately motivated the JT-60U .

Comparison with TFTR

In the late , the original JT-60 achieved significant advances in plasma , with measurements reaching up to 10 keV in discharges using high-power neutral beam heating in plasmas. These results were obtained through Rutherford scattering diagnostics with a beam, which provided profiles during neutral beam injection at energies of 40-75 keV. The TFTR tokamak, operating concurrently, utilized deuterium plasmas initially, with ion temperatures verified through neutron emission spectroscopy. Early experiments on TFTR demonstrated central ion temperatures around 10-20 keV, with diagnostics benefiting from neutron yields for validation. This method contrasted with JT-60's reliance on charge-exchange and scattering techniques in D-D plasmas, where fusion rates are lower and inferred indirectly. Regarding the fusion triple product (n_i τ_E T_i), JT-60 reported a value of 1.2 × 10^{20} m^{-3} s keV in 1990 during advanced experiments with profile control and pellet fueling, enhancing confinement in high-β_p H-mode plasmas. This was calculated from measured central density, energy confinement time, and profiles. TFTR's operations in the late yielded comparable triple products around 10^{20} m^{-3} s keV in supershot configurations, corroborated by diagnostics. The difference in fuel types highlighted challenges in direct comparisons, as D-D rates on JT-60 required scaling to equivalent D-T performance. International reviews in the late and early , including IAEA Fusion Energy Conferences, recognized the complementary roles of JT-60 and TFTR in advancing physics, with JT-60's contributions to high-β operations and current drive validated alongside TFTR's direct measurements.

JT-60U Upgrade

Motivations and modifications

The upgrade to JT-60U was driven by the need to extend plasma pulse lengths beyond the original JT-60's limitations, which were constrained by the air-core transformer's volt-second capacity of approximately 5 V·s, and to enable studies of non-inductive current drive for steady-state operation. These changes were essential to achieve higher plasma performance, such as increased current up to 6 MA and volume to 100 m³, to provide physics data for design and advanced reactor concepts. The retrofit, carried out from 1989 to 1991, replaced the vacuum vessel and poloidal field coils with larger versions to accommodate plasma operations and enhance confinement. A new lower X-point divertor was installed for improved heat removal and particle control, while the first wall was converted from to carbon tiles to minimize impurities. Neutral beam injectors were upgraded to deliver up to 30 MW of power, supporting higher heating and current drive capabilities. The was increased from 2.8 to 3.1 through geometry adjustments, facilitating better . Wall conditioning was enhanced with boronization techniques to reduce oxygen and metal impurities, achieving lower effective charge Z_eff values. Operations resumed in March 1991 following the modifications.

Operational history

JT-60U commenced operations in 1991 following its upgrade from the original JT-60 configuration, marking the start of a 15-year period of intensive experimental campaigns aimed at advancing plasma performance. Initial runs focused on establishing baseline operations and exploring high-current plasma regimes, with full experimental activities continuing through 2008. A pivotal phase began in 1996 with the discovery of reversed shear plasmas, where negative magnetic shear in the plasma core led to the formation of internal transport barriers, enabling enhanced confinement. This breakthrough initiated high-performance campaigns that peaked from 1996 to 2003, during which researchers systematically developed advanced operational modes, including the exploration of steady-state scenarios. In these efforts, steady-state advanced plasmas were sustained for durations up to 28 seconds, demonstrating feasibility for prolonged non-inductive current drive. Operational challenges emerged due to the intense conditions of high-power heating, particularly wall damage from energetic particles and edge-localized modes, which necessitated periodic maintenance downtimes. In 2008, significant wall damage was identified, prompting repairs that briefly interrupted experiments before resumption for final runs later that year. These late-stage experiments, conducted in 2008, prioritized wrapping up key investigations while preparing for facility transition. Operations concluded in August 2008, after which decommissioning activities commenced in 2009 to facilitate the conversion to JT-60SA, involving the systematic dismantling of radioactive components and infrastructure modifications.

Key scientific results

JT-60U experiments pioneered the study of reversed shear plasmas, first observed in 1996, where a safety factor profile with an off-axis minimum led to the formation of internal transport barriers (ITBs) that enhanced confinement significantly beyond standard H-mode levels. These ITBs suppressed anomalous transport in the core, achieving high confinement times while maintaining stability, and represented a key step toward advanced regimes for steady-state operation. In 1998, JT-60U achieved a deuterium-tritium equivalent fusion gain factor of QDTeq=1.25Q_{\mathrm{DT}}^{\mathrm{eq}} = 1.25 in a reversed shear discharge at a plasma current of 2.6 MA, marking the highest equivalent performance in a at the time and demonstrating reactor-relevant thermonuclear conditions. This result was accompanied by a record fusion of nD(0)τETi(0)=1.5×1021 m−3 s keVn_{\mathrm{D}}(0) \tau_E T_i(0) = 1.5 \times 10^{21} \, \mathrm{m^{-3} \, s \, keV}, obtained in high-βp\beta_p H-mode plasmas with central ion temperatures up to 45 keV, underscoring the device's capability to reach ignition-relevant parameters. Non-inductive operations were advanced through high bootstrap current fractions, reaching up to 75% of the total plasma current in reversed shear plasmas, sustained for 7.4 s with nearly full current drive (fCD>90%f_{\mathrm{CD}} > 90\%) at normalized beta βN≈2.4\beta_N \approx 2.4. Long-pulse ELMy H-mode discharges were extended to durations of approximately 30 s at plasma currents up to 1.4 MA, enabling studies of steady-state compatibility with high confinement and heat exhaust. JT-60U contributed critically to design by validating edge-localized mode () control techniques, such as the use of grassy ELMs that reduced energy losses to 0.4–1% of energy, mitigating divertor heat fluxes below 10% of Type I ELM levels through optimized shaping and heating. For disruption mitigation, experiments demonstrated that puffing effectively reduced halo currents (with toroidal peaking factor ×Ih/Ip<0.52\times I_h / I_p < 0.52) and suppressed runaway electrons, while massive gas injection lowered divertor heat loads, informing 's mitigation strategies.

JT-60SA

Project origins and collaboration

The JT-60SA project originated as a key component of the Broader Approach agreement, a collaborative framework established between the European Union (EU) and Japan to accelerate fusion research and complement the ITER experiment by focusing on steady-state plasma operations and advanced tokamak scenarios. The agreement, which includes JT-60SA as its Satellite Tokamak Programme, was initialled in November 2006 and formally signed on February 5, 2007, entering into force on June 1, 2007. This initiative built upon Japan's existing JT-60 facilities at the Naka Fusion Institute, transforming the JT-60U into a fully superconducting tokamak to address physics challenges relevant to ITER and future demonstration reactors (DEMO). The project was approved under the Broader Approach to enhance international cooperation beyond ITER, with operations planned to support long-pulse, high-performance plasmas. The collaboration between Japan and the EU is structured as an in-kind contribution model, with Japan (via the National Institutes for Quantum Science and Technology, QST, formerly Japan Atomic Energy Agency) responsible for approximately 50% of the components, including the poloidal field coils, vacuum vessel, in-vessel components, and the integration building, while the EU (through Fusion for Energy, F4E, and voluntary contributors such as CEA, ENEA, and KIT) provides the other half, encompassing the toroidal field coils, cryostat, power supplies, and high-temperature superconducting current leads. The total construction cost for JT-60SA is estimated at around €560 million (in current values), shared equitably under the Broader Approach and Japan's national fusion program, emphasizing shared expertise in superconducting technologies and plasma control. This partnership ensures joint operation and scientific exploitation by a unified EU-Japan team, fostering knowledge exchange on fusion engineering and physics. Key goals of JT-60SA include achieving plasma currents of up to 5.5 MA in baseline scenarios (with advanced scenarios targeting 7.5 MA), sustaining pulses for 60–100 seconds, and operating at high normalized beta (β_N > 4) to explore stable, high-efficiency plasmas for DEMO-relevant conditions. These objectives aim to validate steady-state operation, current drive techniques, and power exhaust solutions that cannot fully address due to its pulse-length limitations. The design phase spanned 2007–2010, focusing on optimizing the systems and vessel configuration, with commencing in 2013 following detailed reviews.

Assembly and technical challenges

The assembly of JT-60SA began in 2013 following the of the original JT-60U components, with the vacuum vessel sectors progressively installed around the base. Nine 40-degree sectors, totaling 340 degrees, were welded together by late , allowing for the insertion of the 18 toroidal field (TF) coils. The final 20-degree vacuum vessel sector, complete with a TF coil and thermal shield, was welded in April 2018, achieving full closure by September 2018 after precise adjustments to ensure integrity. This closure marked a critical , enclosing the vessel in a double-walled, water-cooled designed to withstand plasma-facing conditions. The TF coils, manufactured primarily in Europe by contributors including ASG Superconductors and , were air-shipped to for integration, with the last two coils arriving in early 2018 to address minor supply chain delays in production. All 18 D-shaped NbTi superconducting TF coils—each weighing approximately 310 tonnes and measuring 7 meters tall by 4.5 meters wide—were installed between 2016 and May 2018 using a specialized rotary crane system for precise positioning around the . These coils, tested individually at full current (25.7 kA) and cryogenic temperatures prior to assembly, generate a peak of 5.65 T at the conductor to confine the plasma toroidally. The , through Fusion for Energy, also supplied the , a 13-meter-diameter enclosure that supports the entire magnet system and maintains superconducting conditions. Installation of the poloidal field (PF) coils proceeded from late 2018 into 2020, building on the completed . The six equilibrium field (EF) coils—three lower units installed in 2016 and three upper units positioned post-TF assembly—along with the four-module central solenoid (CS), were stacked and secured to shape and position the plasma. The CS, the heaviest component at 1,000 tonnes and 12 meters tall, was inserted into the center in May 2019 using an , with final integration completed by December 2019; all 10 PF coils (EF and CS combined) underwent cold testing before full assembly. This phase culminated in March 2020 with the closure of the top lid, finalizing the magnet system's physical construction despite minor disruptions from the emerging , which affected on-site coordination but did not derail the schedule. Key technical challenges during assembly centered on achieving millimeter-level precision in coil alignment to minimize magnetic error fields below 10^{-4}, essential for stable plasma confinement. Adjustable support structures and tracking systems were employed to position the massive TF and PF coils within tolerances of a few millimeters, compensating for the flexibility of unassembled vacuum vessel sectors through pre-assembly trials. Integrating the cryogenic system posed another hurdle, requiring seamless connections for supercritical circulation at 4.4 to cool the NbTi conductors while managing thermal shields and avoiding quench risks; this involved extensive testing of feeders and to ensure uniform cooling across the 3,500-tonne magnet assembly. Supply chain logistics for superconducting materials, including the need for expedited air transport of TF coils, further tested project timelines but were resolved through international collaboration under the EU-Japan Broader Approach agreement.

Commissioning and first plasma

The integrated commissioning tests for JT-60SA began in October 2020, focusing on the cooldown of cryogenic components and superconducting coils to achieve operational temperatures. By January 2021, the first energization of these coils was successfully performed, initiating systematic checks of the magnet systems under controlled conditions. A significant setback occurred on March 9, 2021, during high-voltage testing of the equilibrium field (EF) coils at approximately 5 kV. An insulation failure at the terminal joints of the EF1 coil triggered an arc, creating a short-circuited loop that resulted in a discharge lasting 1.4 seconds and dissipating around 60 kJ of energy. This incident caused damage to the joint shells and a subsequent helium leak into the vessel, though the superconducting windings remained intact. Extensive repairs, including insulation reinforcement and joint redesign, were carried out collaboratively by Japanese and European teams, with work completed by late 2022; integrated commissioning resumed in May 2023 after thorough risk assessments and verification tests. First plasma was achieved on October 23, 2023, marking a key milestone with an initial plasma current of 130 kA sustained in a toroidal magnetic field of 2.3 T for several seconds. This low-power operation validated the core functionality, including plasma initiation and basic confinement. In 2024, further tests energized the central up to 10 kA, confirming its role in inductive plasma current drive without anomalies. During these initial phases, key diagnostics were commissioned and validated, including soft imaging diagnostics that measured electron temperature profiles with high , and magnetic measurement arrays that accurately reconstructed plasma equilibrium and position. These systems provided essential data for refining control algorithms and ensuring diagnostic reliability ahead of higher-performance operations.

Current operations and upgrades

Following the achievement of first plasma in October 2023, JT-60SA conducted its initial plasma campaign from 2023 to 2024, focusing on basic plasma formation and control. Integrated commissioning activities continued into 2025, with upgrades preparing the device for enhanced performance. Full scientific operations are scheduled to commence in 2026, enabling advanced experimentation in support of . Recent results from the first plasma campaign have emphasized plasma control studies, including equilibrium reconstruction and control using both Japanese and European codes. These efforts demonstrated access to high-beta plasmas, with normalized beta values approaching ITER-relevant regimes during H-mode operations. Such studies, detailed in a May 2025 publication, highlight progress in disruption avoidance and steady-state sustainment for durations up to 100 seconds at plasma currents of 5.5 MA. Ongoing upgrades in 2025 include the installation of new diagnostic systems to improve plasma measurement capabilities. The edge diagnostic, a joint Europe-Japan effort, had its structural components installed in September 2025 to enable detailed edge plasma profiling during upcoming campaigns. In November 2025, the imaging crystal spectrometer (XICS), provided by the U.S. Department of Energy's Princeton Plasma Physics , was delivered for installation in early 2026 to enhance ion temperature and flow diagnostics. Additionally, a cooperation agreement was signed on October 11, 2025, between the JT-60SA project and to advance research on energetic particle behavior and fusion efficiency. These enhancements, along with increased heating power and divertor modifications, aim to boost overall performance for the 2026 return to operations. Looking ahead, JT-60SA operates under a 20-year research plan coordinated between and , emphasizing joint experiments to mitigate risks and extrapolate to DEMO reactor designs. This includes integrated studies on power and particle exhaust, high-beta steady-state scenarios, and metal-wall operations starting in the integrated research phase II. Through the Broader Approach agreement, these activities will directly support 's operational optimization and fusion energy development.

Specifications

Plasma and vessel parameters

The original JT-60 featured a vacuum vessel supporting plasma configurations with a major radius of 3.04 m, a minor radius of 1.12 m, and a plasma volume of approximately 90 m³, enabling plasma currents up to 3.2 MA. These dimensions allowed for initial studies of high-current plasmas in a circular cross-section , with the vessel designed to accommodate toroidal fields up to 4 T. The JT-60U upgrade retained a similar but introduced modifications for elongated plasma shapes, achieving an elongation factor κ of up to 1.7 while maintaining a plasma volume of about 90 m³. This evolution facilitated advanced configurations with plasma currents up to 5 MA, emphasizing improved stability and confinement in non-circular plasmas without significantly altering the overall vessel scale. In contrast, JT-60SA represents a major redesign with a superconducting vacuum vessel optimized for long-pulse operations, featuring a major radius of 3.15 m, a minor radius of 1.2 m, and a substantially larger plasma volume of approximately 130 m³. The baseline plasma current is 5.5 MA, supported by an integrated divertor system capable of handling high fluxes exceeding 10 MW/, enabling sustained high-performance plasmas in a lower geometry. In operations as of 2024, it has achieved a plasma volume of 160 m³, setting a for the largest plasma volume. Across JT-60 iterations, key include aspect ratios ranging from 2.6 to 3.1, normalized beta values up to 4.5 for advanced steady-state scenarios, and lengths extending to 100 s in JT-60SA, reflecting progressive enhancements in confinement and stability.
ParameterOriginal JT-60JT-60UJT-60SA
Major radius (m)3.043.43.15
Minor radius (m)1.121.0 (elongated)1.2
Plasma (m³)~90~90~130 (; achieved 160)
Plasma current Ip (MA)3.2 (max)5.0 (max)5.5 (baseline)
Elongation κ1.0 (circular)Up to 1.7Up to 1.95
~2.7~3.42.6–3.1
length (s)~1015100

Magnet and power systems

The magnet and power systems of JT-60U relied on resistive coils for magnetic confinement, with the toroidal field (TF) system powered by a rated at 215 MVA to achieve a maximum toroidal field of 4.2 T. The poloidal field (PF) system featured coils with currents up to 120 kA in the ohmic heating (OH) coil and an air-core OH , enabling plasma shaping and currents up to 3.2 MA for pulses of approximately 15 s. The overall power infrastructure supported short-pulse operations, with PF power supplies modified to increase stored capacity from 100 MJ to 150 MJ during upgrades. In contrast, JT-60SA incorporates a fully system designed for long-pulse, steady-state plasmas, marking a shift from resistive losses to zero-loss operation. The TF system comprises 18 D-shaped NbTi coils, each with 72 turns carrying 25.7 kA for a per-coil amp-turns of 1.85 MA (total ~33 MA across all coils) and a peak field of 5.65 T in the conductor, producing a 2.25 T field at the plasma center. The central (CS) consists of four Nb₃Sn modules, each with 549 turns at 20 kA (11.1 MA amp-turns per module), achieving a peak field of 8.9 T. The six superconducting NbTi PF coils support currents of 20 kA, with peak fields ranging from 4.8 T to 6.2 T, enabling advanced plasma shaping such as lower single-null configurations. All coils use cable-in-conduit conductors cooled by supercritical at 4.4–4.5 K, with a cryogenic system providing 9 kW equivalent refrigeration capacity at 4.5 K. The JT-60SA power systems have been upgraded for high-capacity, long-duration pulses, with a total facility power of approximately 500 MW supporting , heating, and diagnostics. Magnet power supplies include high-temperature superconducting (HTS) current leads rated for 25.7 kA (TF) and 20 kA (CS/PF), connected to 20 kV converters capable of 100 kA output for sustained operation up to 100 s. A motor-generator system rated at 400 MVA minimizes grid fluctuations, while poloidal field circuits use bidirectional 12 kA converters for precise control. This evolution from JT-60U's resistive setup to JT-60SA's superconducting infrastructure reduces energy losses and enables extended high-beta plasma experiments critical for complementarity.

Heating and diagnostics

The heating systems of the original JT-60 tokamak primarily utilized neutral beam injection (NBI) with up to 20 MW of power from beams for core plasma heating and current drive, supplemented by electron cyclotron resonance frequency (ECRF) heating at 4 MW to enable localized heating and off-axis current drive. During the upgrade to JT-60U, a lower hybrid current drive (LHCD) system was added, providing up to 7 MW of power through a multijunction launcher to enhance non-inductive current drive and . In the JT-60SA configuration, the heating capabilities have been significantly expanded for steady-state operations, with NBI increased to 34 MW (including both positive-ion and negative-ion sources) and ECRF to 7 MW at 140 GHz, achieving a total auxiliary heating and current drive power of 41 MW for durations up to 100 seconds. Current drive in JT-60 evolved to support advanced scenarios, particularly through LH waves in JT-60U, which enabled off-axis current deposition to optimize the safety factor profile and complement the bootstrap current for non-inductive operation. This approach facilitated enhanced bootstrap current fractions by tailoring the current profile, reducing the need for inductive drive and improving confinement in reversed shear plasmas. In JT-60SA, the integrated NBI and ECRF systems continue to support these functions, with flexible beam geometries allowing precise control of current profiles for high-beta steady-state discharges. Diagnostics on JT-60 provided essential measurements for plasma characterization, including Thomson scattering systems using YAG and ruby lasers to determine electron temperature (Te) and ion temperature (Ti) profiles across the plasma core and edge. Neutron detectors, such as fission chambers, monitored fusion reaction rates and neutron flux to assess plasma performance and neutronics, while bolometers measured radiated power losses to evaluate energy balance and impurity content. The evolution of these systems progressed from basic profile measurements in the original JT-60 to advanced configurations in JT-60U and JT-60SA, incorporating real-time feedback capabilities for active control of plasma parameters like current profiles and stability. For JT-60SA, upcoming enhancements include an X-ray imaging crystal spectrometer (XICS) scheduled for 2026 to measure ion temperature and rotation profiles, and an edge Thomson scattering diagnostic installed in 2025 for high-resolution pedestal measurements.

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