National Policy of Future Nuclear Fusion Research and Development
- Supplementary Volume -
- Glossary -
26 October 2005
Atomic Energy Commission
Advisory Committee on Nuclear Fusion
- Self-ignition condition *1
Consider plasma with deuterium (D) and tritium (T) as fuel and heat it externally. With heating input, its temperature increases continuously and will reach at a point where the heating energy due to alpha particles produced by fusion reaction is balanced by the energy lost from plasma. This condition is the self-ignition condition. Under this condition, it is expected that the fusion reaction can be maintained even if the continuous external heating input is cut off. In order to reach this condition, confinement time of plasma energy, plasma temperature, and plasma density must satisfy a certain relation. This condition requires higher confinement performance than that of break-even condition.
- Long-duration burning *2
A state where DT reaction etc. is maintained continuously for long duration in a fusion reactor with D and T as its fuel. Here, by regarding the fusion reaction as heat source, it is compared to the burning in chemical reaction.
- International Thermonuclear Experimental Reactor (ITER) project *3
A tokamak-type fusion experimental reactor project, aiming at demonstrating scientific and technological feasibility of nuclear fusion, by sustaining controlled fusion burning plasma and long-duration burning. The project has been promoted since 1992 as international cooperation among Japan, US, Europe, and Russia, and conducted 9 years of engineering design and technological development of main components. At present, formal intergovernmental negotiations toward its construction are undertaken by Japan, Europe, Russia, US, China, and Korea.
1.1 Role of Fusion Energy on Solving Energy and Environmental Problems
- Nuclear fusion energy *4
Energy generated by nuclear fusion reaction. Energy produced from the fusion reaction of one gram of deuterium (D) and tritium (T) fuel corresponds to the amount of heat produced by burning a tankerload of oil (about 8 tons).
1.2 Significance and Necessity of Fusion R&D in Nuclear Energy Policy
- Tokamak type system *5
A toroid-shaped confinement system in which the plasma is confined by magnetic fields. The main magnetic field is toroidal magnetic field in toroidal direction. However, only with this, plasma cannot be confined. In order to balance the plasma pressure with magnetic force and get equilibrium, poloidal magnetic field is also needed. Poloidal magnetic field can be produced by generating current in the plasma in the toroidal direction. Plasma current also plays the role of plasma heating through Ohmfs law. This system was devised at Kurchatov Institute of the former Soviet Union, and because of its superior confinement property, this type of plasma experimental devices have been constructed and studied in many institutes all over the world.
- Helical type system *6
A toroid-shaped confinement system in which the plasma is confined by magnetic fields, similar to tokamak system. But in contrast to tokamak, the poloidal magnetic field necessary for plasma confinement is produced not by plasma current but by external coils. For external coils, spirally twisted coils (helical coils) or more complicatedly shaped modular coils are used. A system that confines plasma with these methods is called helical system.
- Laser type system *7
A system in which a fuel pellet of a few millimeters in diameter is imploded (adiabatic compression) isotropically by laser to produce instantaneously ultra-high-density high-temperature plasma and initiate fusion reaction, is called laser type system.
- Plasma *8
As temperature increases, the state of matter generally transits from solid to liquid, and to gas. And if the temperature further increases, electrons orbiting around nucleus are stripped to form a state where atoms are dissociated to positively charged ions and negatively charged electrons (ionization), both of which exhibiting fast random motion. This state is called plasma. Nuclear fusion deals with ultra-high temperature plasmas up to several hundreds million degrees. Plasma exists widely in nature such as lightening and aurora. For a familiar example, we can find plasma produced by discharge in rarefied gas such as in a fluorescent lamp.
- Break-even test facility, JT-60 *9
The abbreviation of the break-even test facility, JAERI Tokamak-60, and is one of the world largest tokamak devices operated at Japan Atomic Energy Research Institute, Naka Establishment (major radius R = 3.4 m, minor radius a = 1.0 m, toroidal magnetic field Bt = 4.0 T, plasma current Ip = 3.0 MA). Along with TFTR (operation completed) in US and JET in EU, these were called Three Large Tokamaks. The world highest temperature of 520 million degrees achieved in JT-60 is registered in the Guinness Book of World Records.
- Break-even plasma condition *10
Define the energy gain value Q by the ratio of the energy produced by nuclear fusion reaction to the energy input to plasma. Then the break-even plasma condition refers to plasma that would satisfy Q = 1 if the plasma ions consisted of the same numbers of deuterium and tritium. The break-even condition is given in terms of the product of plasma temperature, plasma density, and energy confinement time. In JT-60 (Japan) and JET (EU), plasma parameters exceeding break-even condition have been achieved.
- Fusion core plasma *11
Generic term for plasma produced in researches aiming at fusion reactor.
- Academic research *12
Research on academics, which is the generic term for studies and arts that refer to knowledge and method systematized based on certain theories. But here, we refer to the academic research related to nuclear fusion as one of the fields in natural science. In aiming at the realization of nuclear fusion, academic research advances by putting a high priority on the establishment of systematized knowledge.
- Developmental research *13
Research and development conducted systematically for practical purpose. Here, a higher priority is put on the selection of important issues for development/performance goal, analysis of methods to solve these issues, and achievement of goal by the carrying out of research.
- Fusion technology *14
The research fields indispensable for the technological development of nuclear fusion reactor (blanket technology, superconducting magnet technology, reactor structure and remote maintenance technique, heating and current drive equipment technology, plasma facing component technology, tritium fueling, pumping and circulating processing technology, diagnostic and control device technique, nuclear fusion neutronics technology, nuclear fusion material technology, safety engineering technology, etc.).
- ITER physics R&D *15
ITER physics R&D is an activity in which result of research and development conducted on a voluntary basis in each participating party is provided to the Engineering Design Activity (EDA) of ITER. It started from 1994. Under the ITER Physics Committee, which consists of ITER Director General and the Committee members etc. designated by each party, seven Expert Groups (Transport and Internal Barrier Physics; Confinement Database and Modeling; Edge and Pedestal Physics; Scrape-off Layer and Divertor Physics; MHD, Disruptions and Control; Energetic Particles; Heating, and Steady State Operation; Diagnostics) were organized. In this activity, identification of physics issues, and examination and evaluation of research results provided by each party were conducted and reflected to the ITER project. ITER Physics R&D was completed in July 2001, and was succeeded by the International Tokamak Physics Activity (ITPA).
- ITER technology R&D *16
Research and development for validating the engineering technology of ITER in the ITER design activity. Seven major technology R&D items (development of ITER center solenoid model coil, development of ITER toroidal magnetic field coil, development of large-scale vacuum vessel, development of blanket, development of divertor, development of remote maintenance technology for blanket, development of remote maintenance technology for divertor) were completed in 2001.
- Breeding and power-generating blankets *17
For a fusion reactor using the reaction of deuterium and tritium, the fuels are deuterium and tritium. But since tritium is not a naturally occurring substance, we have to make it artificially. Therefore a method is considered, in which the structure surrounding the plasma, where fusion reaction is occurring, is packed with lithium compound and set in place, and by utilizing the neutron produced in fusion reaction, lithium atom is transmuted to tritium via nuclear reaction. The lithium compound for this purpose and the structure that contains it is called tritium-breeding blanket. Likewise, a blanket for converting neutronfs kinetic energy to heat and using this heat to generate power is called power-generating blanket.
2.2 Experimental Reactor Program
- Steady-state fusion reactor *18
A fusion reactor that maintains fusion reaction continuously. In contrast, a fusion reactor that repeats the reaction for certain period and operates in an intermittent manner is called a pulsed reactor.
- Major radius, Minor radius, Plasma current, Q value *19
Major radius is the radius in the long-way-around direction of doughnut-shaped plasma, that is the distance from the center of the device to the cross-sectional center of the doughnut-shaped plasma. Minor radius is the radius in the short-way-around direction of doughnut-shaped plasma, that is the radius of plasma cross section. The current that flows along the toroidal direction (torus direction) is called plasma current. And the ratio of the output from fusion reaction to the heating input necessary to maintain the plasma state (i.e. heat loss from the plasma) is called Q value (Q = output/input).
- Fusion burning *20
The state where DD or DT fusion reaction occurs is called fusion burning. Even in the case where break-even condition or self-ignition condition is not satisfied, we call it a fusion burning plasma, if fusion reaction is occurring. In the sun, four protons, which are the nuclei of hydrogens, are fused to yield the nucleus of helium. Thus fusion reaction is occurring and it is fusion burning.
- Alpha particle *21
In DT fusion reaction, alpha particle (nucleus of helium) with the energy of 3.5 MeV is produced and used for the heating of fusion plasma of several tens of keV. Since alpha particle has electrical charge, it is confined in magnetic field and its energy is gradually transferred to plasma particles by collision. As a consequence, the plasma is heated.
- Superconducting magnet *22
An electrical magnet (coil) using superconductor. Since superconductor has zero resistivity, the operation of superconducting magnet, unlike normal conductor, does not yield joule loss (heat). Also, since superconductor can conduct large current with a small cross section, it is suitable for making a compact electrical magnet.
- Vacuum vessel *23
Vacuum vessel is a toroidal hollow metal container to confine plasma.
- Plasma facing components *24
Generic term for components that are set to the place directly facing to plasma. They include components such as first wall, divertor, and limiter, and constitute the wall surrounding the plasma.
- Remote maintenance equipments *25
Since the structural objects in vacuum vessel (blanket, divertor, etc.) are radioactivated by the neutrons produced in fusion reaction, in case that they are damaged, they must be repaired or replaced by remote operation. Remote maintenance equipments is a generic term for various equipments used for the above remote maintenance.
2.3 Fusion Plasma Research
- Energy confinement time *26
The efficiency of how little heating power is required to maintain high-temperature high-density plasma. It is equal to the time scale for plasma to cools down after the heating is stopped. A longer energy confinement time means that the plasma is less apt to cool down.
- Stability of plasma *27
The tokamak plasma is confined by a magnetic field. When the pressure of the plasma and the pressure of the magnetic field become unbalanced or when the plasma current becomes too large for a given toroidal field, the confinement performance deteriorates or in the worst case the plasma disappears. This is called instability of the plasma. Stable sustainment of high-pressure plasma requires the optimization of spatial profiles of the plasma pressure and the plasma current density.
- Non-inductive current drive *28
In a tokamak, plasma is maintained by driving current in the plasma, which corresponds to the secondary winding in the principle of transformer (inductive method). But with this method, it is impossible to drive current for a long period. In order to operate a tokamak type fusion reactor in a steady-state manner, it is necessary to drive plasma current by non-inductive methods (non-inductive current drive). The non-inductive currents include a naturally generated current in proportion to the pressure gradient of the plasma (bootstrap current) and an externally driven current (external current drive) by using radio frequency waves (electron current by the principle of surfing) and neutral beam injection (current by ions).
- Heat and particle control *29
In a fusion reactor, ultrahigh-temperature plasma needs to be maintained for a long time. Since the wall surrounding plasma is subjected to extremely high heat flux and plasma particle flux, atoms composing the wall material are ejected into the plasma by various interactions (sublimation, melting, sputtering, thermal desorption, etc.). If the ejected particles enter the high-temperature plasma, they lower the plasma temperature and then decrease the fusion power. Therefore, it is necessary to reduce the number of ejected particles by decreasing the heat flux to the wall or enhance the shielding effect of the peripheral plasma against the penetration of ejected particles. Namely, it is necessary to control these heat flux and the particle flux effectively in order to maintain the state where sufficient fusion reaction occurs for a long time.
- International Tokamak Physics Activity (ITPA) *30
An international research activity succeeding the ITER physics R&D that had been conducted during ITER/EDA, and started from 2001 so the researchers from the four parties, EU, USA, Russia and Japan, as to clarify the physics of tokamak plasma and assess the performance of fusion burning plasma in ITER etc. It consists of a Coordinating Committee and seven Topical Groups: Confinement Database and Modeling; Transport Physics; Divertor and SOL; Pedestal and Edge; Steady State Operation; MHD; Diagnostics.
- JFT-2M *31
This is the acronym of advanced tokamak development test device, JAERI Fusion Torus-2M. It was a medium-sized tokamak device (R = 1.3 m, a = 0.35 x 0.53 m, Bt = 2.2 T, Ip = 500 kA) operated (operation started in April 1982, completed in March 2004) at Japan Atomic Energy Research Institute, Tokai Research Establishment (division belongs to Naka Fusion Research Establishment). Researches that took advantage of its flexibility, such as H-mode physics, edge plasma control using closed divertor, MHD mode suppression, and compatibility test of ferritic steel were carried out and their results contributed to the divertor design and MHD mode suppression/control in JT-60 and to ITER physics R&D.
- H-mode (high confinement mode) *32
H-mode is an improved confinement state in which the energy confinement time is roughly doubled, accompanying steep slopes in the temperature and density profiles near the plasma surface, during the heating with neutral beams or radio frequency waves. It was first discovered on the ASDEX tokamak (Germany) that had divertor, followed by researches in world tokamak devices, and it is now regarded as a standard operation mode of ITER.
- Scaling law *33
This refers to a relational expression, which shows how the energy confinement time of plasma depends on the dimensions of a tokamak device (major radius, minor radius, toroidal field, heating power, etc.) and parameters of plasma (density, plasma current, etc.). Empirical scaling laws obtained from various experimental data are often employed. Different scaling laws have been obtained individually for confinement of ohmic-heated plasmas, standard L-mode plasmas and high-performance H-mode plasmas. Through Conceptual Design Activity, Engineering Design Activity, and International Tokamak Physics Activity, the plasma confinement data of the tokamaks of all over the world are collected, and the confinement scaling law with high prediction accuracy has been deduced for the ITER design.
- Disruption *34
A phenomenon, in which the structure of the magnetic field within the plasma changes, followed by a rapid decrease in plasma current and the disappearance of the plasma. It may arise, for instance, when a large amount of impurity enters plasma.
- Lower hybrid wave *35
A wave in plasma with a frequency range between the electron cyclotron resonance and the ion cyclotron resonance. The frequency is in the range of several GHz for a tokamak-type fusion reactor. Though its current drive efficiency is high, the current drive is difficult in the central region of high-temperature high-density plasma and the heat load on the antenna is severe. Hence, it is regarded as an option in the upgrade of the heating / current-drive system in ITER.
- Negative ion based ion source *36
A device, which generates a high-energy negative ion beam. Plasma is generated by arc discharge or radio-frequency wave. Then negative ions are generated using reactions in the plasma and/or reactions on the electrode surface attaching the plasma. Generated negative ions are usually drawn and accelerated by an electric field applied between multiple pieces of electrodes to form a negative ion beam. In a high-energy neutral beam injector, the conversion efficiency to neutral particles can be drastically improved by using a negative ion beam instead of a positive ion beam. In JT-60U, a high energy beam over 400 kV was injected into a tokamak plasma, and improvement of confinement performance and current drive performance was demonstrated. It is also used in a device to manufacture a thin film of semiconductor as well as in a heating system for nuclear fusion.
- Neutral beam injector *37
Neutral beam injector (NBI) is a device to generate high-energy ion beam (primary particles) in ion source, convert it to fast neutral particles (atoms) without electric charge, and inject them into a fusion plasma. Both positive ions and negative ions are used as primary particles. A high-energy beam with a MeV class is required for heating and current drive of a high-density plasma like that of ITER. In such a high-energy region, it is indispensable to use negative ions as primary particles that have a high conversion efficiency to neutral particles.
- Electron cyclotron wave injection system *38
The device for heating electrons and generating a current in a plasma by injecting an electromagnetic wave at the rotation frequency of electron (electron cyclotron frequency) or near at its harmonics. It consists mainly of a high-power RF source (gyrotron), which generates an electromagnetic wave in the electron cyclotron range of frequency (100 GHz band in a tokamak fusion reactor), a transmission system (waveguide), which transmits the electromagnetic wave, and a coupling system (launcher), which launches the electromagnetic wave into a plasma. Owing to its high frequency and short wavelength in a millimeter range, it has a merit that wave can be injected using mirrors as couplers from a long distance like a laser beam.
- Helium ash *39
Low-temperature helium ions that are generated in the nuclear reaction of deuterium (D) and tritium (T) as alpha particles (helium nuclei) with energy of 3.5MeV, and gradually lose energy through the collision with plasma particles. It is called ash because it is a combustion product in a fusion reactor. It is necessary to exhaust helium ash efficiently since the helium ash remaining in a plasma dilutes the fuel ion density and lowers the fusion output.
- Plasma-surface interaction *40
In a tokamak device, plasma particles escaping from the main plasma impact on the vacuum vessel wall (the first wall). As a result, the wall surface is eroded by sputtering, chemical reactions, and evaporation. Furthermore, the particles injected into the wall surface are ejected again to the side of plasma surface through reflection, capture and diffusion. Such plasma-surface interactions are an important research subject of the nuclear fusion development since they are closely related to impurity and density control of plasma, wall erosion, fuel retention in the wall, and so on.
- Tritium *41
A radioisotope of hydrogen with nucleus consisting of one proton and two neutrons. It has radioactive half-life of 12.3 years, emits beta ray whose maximum energy is 18.6 keV and average is 5.7 keV, and disintegrates into 3He. In the nature, it is created by reaction between cosmic rays and atmosphere constituent elements, and the production amount is evaluated to be about 160-200 g/year.
- JET (Joint European Torus) *42
A large tokamak device (R = 3 m, a = 1.25 m, Bt = 3.5 T, Ip = 3.0 MA) run by EU. It is situated at Culham Science Center, in the United Kingdom. Along with JT-60 of Japan, TFTR of the United States, it is called one of the Three Large Tokamaks. The experiment was started in June 1983. After that improvement and reinforcement of various parts of the device were carried out. And in 1991, a DT-equivalent fusion gain (Q-value) of 1.1 was achieved in a deuterium discharge. Preliminary DT discharges, in which about 10% of tritium was mixed in deuterium, were carried out in 1991 for the first time in the world. In 1997, D-T fusion power of 16 MW was realized, though the duration period was as short as no more than 1 second. Afterwards, it is continuing experiment to improve the physical database for ITER.
- TFTR (Tokamak Fusion Test Reactor) *43
A large tokamak device (R = 2.4 m, a = 0.8 m, Bt = 5.0 T, Ip = 2.2 MA) at Princeton Plasma Physics Laboratory in the United States. The experiment was started at the end of 1982. A DT-equivalent fusion gain of 0.3 was achieved in a deuterium discharge in 1988. After that, it was equipped with a tritium handling facility, and DT discharges with deuterium and tritium ratio of 1:1 were performed since 1993. TFTR achieved a plasma temperature of 5.1 hundred million degrees and a fusion output power of 10 MW in 1994. The operation was completed in April 1997.
- TRIAM-1M *44
A high-toroidal-field tokamak device at Research Institute for Applied Mechanics, Kyushu University. TRIAM-1M is the world first superconducting tokamak with toroidal field coils made of Nb3Sn wire rod (toroidal field is up to 8 T). The experiment was started in 1986. Since 1987, continuous running of the worldfs only Nb3Sn superconductive system for 100 days was carried out every year, which demonstrated durability and stability of the superconducting coil using Nb3Sn. Moreover, it succeeded in maintaining a tokamak plasma for a long time (5 hours and 16 minutes, the world record) only by radio frequency waves (current drive using lower hybrid waves).
- Energy multiplication factor *45
The ratio of the fusion reaction output to the external input supplied directly to a plasma in order to maintain the plasma state. The cases in which this value is unity and infinite are called the break-even condition and the self-ignition condition, respectively. JT-60U achieved this value of 1.25 in June 1998, which is the highest value in the world.
- Negative magnetic shear operation *46
An operation of tokamak where the internal magnetic shear becomes negative in the confinement magnetic field configuration of tokamak plasma. Here, the magnetic shear indicates the extent of twist between the adjacent magnetic surfaces, and is defined by , where q and ρ are the safety factor and the volume averaged minor radius, respectively. In a usual tokamak, the current density profile has a convex shape that has maximum at the center (magnetic axis), and the magnetic shear is positive everywhere in the plasma. When the current profile is changed to a concave shape by some methods, the safety factor q has the minimal value qmin, and the region inside its radius becomes a negative magnetic shear region where the magnetic shear is negative. Such a configuration is called the negative magnetic shear configuration or the reversed magnetic shear configuration. Operation with negative magnetic shear configuration is highly compatible with a large fraction of the spontaneous current (bootstrap current), and is considered to be an operation suitable for a steady-state tokamak reactor.
- DT equivalent energy multiplication factor: QDTeq *47
An energy multiplication factor that would be obtained if half of the fuel deuterium in a DD fusion reaction experiment were replaced by tritium, and by calculating the output power of DT fusion reaction assuming that temperatures and densities were unchanged.
- Spontaneous current (Bootstrap current) *48
A current that is spontaneously driven parallel to a magnetic field due to a radial gradient of the plasma pressure in a torus plasma. A larger spontaneous current leads to a more economical operation because the externally driven current can be reduced. It is one of the crucial factors in steady-state operation of a tokamak reactor. It is assumed that 70-80% of the total plasma current is driven by the spontaneous current in a steady-state tokamak reactor.
- Neutral particle beam *49
Atom (particle) beam injected into a plasma from the outside by a neutral beam injector for plasma heating, current drive for steady-state operation, or plasma diagnostics. In a neutral beam injector, the ion beam is generated by accelerating ions produced in an ion source, and then, since the ion beam is deflected by a strong magnetic field in a tokamak, the beam is converted to a neutral atomic beam by stripping the electric charge before they are injected into a plasma.
- High beta operation *50
The beta value is the ratio of the plasma pressure to the pressure of a magnetic field that confines the plasma. Namely, beta value = (plasma pressure) / (pressure of magnetic field). The higher beta operation means that it is possible to confine a high-pressure plasma with a weaker magnetic field. Since the output power of a fusion reactor is proportional to (square of plasma pressure) x (plasma volume), raising the beta leads to a compact (small volume) reactor. The high beta operation is an essential factor to improve the economical efficiency of a fusion reactor.
- First wall *51
A generic name of the wall that faces a plasma directly. It includes limiters, divertor plates, high heat flux protective walls, and blanket walls, according to functional classification. In a narrow sense, it means the blanket wall whose surface is parallel to a magnetic flux surface. The first wall generally receives a large amount of heat and particle load through the direct interaction with a plasma. Therefore, impurities released from the first wall have no negligible effects on the plasma. The first wall is designed through comprehensive evaluation of heat removal, impurity ejection, particle recycling rate, surface erosion, radiation damage, thermal fatigue, electromagnetic force, and so on.
- Low activation ferritic steel *52
The most promising candidate for a blanket structural material of a fusion demonstration reactor. It was designed on the basis of heat-resistant ferrous materials so as to reduce long-life radioactive waste. F82H, ferritic steel with 8% chromium and 2% tungsten, was developed by JAERI as a leader, in cooperation with universities and others. The performance evaluation is now under way toward an engineering demonstration stage, the last stage before the practical application. Though the influence of its ferromagnetism on plasma confinement magnetic field was a concern, it has been verified that the influence is not a large problem through the compatibility test in the JFT-2M tokamak.
- Ferromagnet *53
A generic name of materials that are magnetized in a magnetic field like iron and nickel. It must be used with care in a plasma confinement device, since it modifies the shape of the confinement magnetic field. Low activation ferritic steel, a promising candidate for a blanket structural material of a fusion demonstration reactor, belongs to ferromagnet.
- Normalized beta value *54
It is shown, by extensive experiments and theoretical calculations, that the upper limit of the beta value in a tokamak plasma is proportional to the plasma current, and is inversely proportional to the toroidal magnetic field and the plasma minor radius. The proportional coefficient is called the normalized beta value, βN (normalized beta value). Raising βN allows realization of a compact and highly efficient tokamak fusion reactor.
- Large Helical Device (LHD) *55
The world largest helical (non-axisymmetric) type experimental device at National Institute for Fusion Science, National Institutes of Natural Sciences, Japan. LHD is the acronym for Large Helical Device. The LHD has twisted (helical) winding coils and adopts the magnetic field configuration originally developed in Japan (Heliotron configuration) in order to confine a plasma. The LHD is equipped with two superconducting helical coils and 3 pairs of superconducting circular (poloidal) coils. The experiment was started in March 1998. The helical type devices are intrinsically superior in controllability. It is said that they are also suitable for steady-state operation required in a future power reactor. Since the magnetic field structure is fundamentally different from that in a tokamak, by conducting research complementary to that on tokamak, a role as a support-device for the promotion of ITER project is expected.
- Fast ignition *56
A technique to achieve ignition and fusion burn by irradiating an ultra high-intense laser to a high-density compressed fuel pellet. The fast ignition gives a high fusion gain with a far smaller laser system than the central ignition method.
- FIREX project *57
A research project conducted by Institute of Laser Engineering, Osaka University, to clarify the scenarios for efficient nuclear fusion ignition and self-burning and to demonstrate the principle of the fast ignition nuclear fusion, by heating the imploded plasma instantaneously with a short pulse ultra high-intense laser.
- CHS (Compact Helical System) *58
A small-scale helical type magnetic confinement device (Compact Helical System) at National Institute for Fusion Science, Japan. Major radius is 1 m, minor radius is 0.2 m, and pitch number (number of the twist) of the helical coils is 8. The CHS supports the LHD experiment through its flexibility as a small device.
- Magnetic field ripple *59
In making doughnut-shaped spiral confinement magnetic field using twisted coils (helical coils), variation of the magnetic field strength arises along the field line, since the magnetic field is stronger just beneath the coils than between the coils. Such a magnetic field structure is called "magnetic field ripple".
- High energy particle *60
A generic name for the particles that have a higher energy than usual thermal particles in a plasma, such as high-energy ions generated by neutral beam injection, runaway electrons generated through acceleration by an electric field, and alpha particles produced by the fusion reaction.
- Heliotron-J *61
A helical type magnetic confinement device at Institute of Advanced Energy, Kyoto University, Japan. It increases the degree of controllability in basic factors of the magnetic field spectrum (toroidal, helical, bumpy) owing to the advanced magnetic confinement configuration (the helical-axis heliotron), allowing development of a new parameter region and flexible experiments in the magnetic field configuration research.
- Inertial confinement fusion *62
Inertial confinement fusion refers to methods to inject large energy instantaneously into solid fuel using devices such as highly intense laser, heavy ion beam or Z pinch, and increase the density and temperature of fuel while it stays due to inertia, and initiate fusion reaction.
- Laser fusion *63
A kind of inertial confinement fusion, in which a highly intense laser is used as a driver.
- GEKKO XII laser system *64
An inertial confinement device using the glass laser at Institute of Laser Engineering, Osaka University. The number of the laser beam lines is 12, the laser energy is 30 kJ (laser wavelength of 1.05 μm), 15 kJ (0.53 μm) and 10 kJ (0.35 μm).
- Implosion *65
In inertial confinement fusion, when a solid fuel particle (pellet) of several millimeters in diameter is uniformly irradiated by a powerful laser beam, or X-ray produced by a laser beam or an ion beam, the energy is absorbed on the surface of the pellet, and heated particles scatter outward. The pellet is compressed inward by their reaction (effect of rocket when it is propelled). This phenomenon is called "implosion".
- Central ignition *66
Irradiated with a large number of powerful lasers, a fuel pellet is compressed to an ultra-high density by the pressure of the plasma produced on the surface (implosion), and then the fusion reaction is ignited with a high-temperature plasma produced at the center of pellet, which is followed by the burning of surrounding fuel.
- NIF (National Ignition Facility) *67
National Ignition Facility, NIF, is an inertial confinement fusion facility of the central ignition type under construction at the Lawrence Livermore National Laboratory (LLNL), US, with planned specification of 1.8 MJ, 0.35 μm, and 192 beams.
- LMJ (Laser Mega Joule) *68
LMJ (Laser Mega Joule), which was approved by Atomic Energy Commission (CEA) of France in 1993, is an inertial confinement fusion facility of the central ignition type under construction, with planned specification of 2 MJ, 0.35 μm, and 240 beams.
- Ultra high-density plasma *69
In the inertial confinement fusion, a spherical shell fuel pellet of a millimeter size is uniformly irradiated with a powerful laser beam. Then high-pressure plasma is produced and accelerates the spherical shell fuel toward the center. The central fuel is compressed (imploded) to produce an ultra-high density state, more than several 100-1000 times higher than that of solid. This is called the ultra-high density plasma.
- Ultra-short pulse laser *70
An ultra-short pulse laser is a laser with a high peak power and a very short duration (typical pulse width is from 1 ps to less than several tens fs, where 1 ps = 1 picosecond = 10-12 second, 1 fs = 1 femtosecond = 10-15 second).
- Peta-watt laser *71
A laser system in which an output power exceeds 1 PW (peta-watt, where 1 P = 1015 = 1000 trillion).
- Reversed field pinch *72
A system in which a doughnut-shaped plasma is formed by applying a strong electric field in toroidal direction and generating a large plasma current compared to the toroidal magnetic field. It is called the reversed field pinch type, since the direction of the magnetic field in the central region is reverse to that in the peripheral region.
- TPE-RX *73
A reversed field pinch (RFP) machine run by National Institute of Advanced Industrial Science and Technology (AIST), Japan. It is one of the largest RFP machines in the world.
- Poloidal current drive *74
Though the reversed field pinch system has an advantage that a large plasma current can be driven, the energy confinement performance is an issue since the confinement magnetic field is formed and maintained by the instability arising in the plasma. Therefore, the research is under way to drive the current by a rotating magnetic field in the poloidal direction, in order to maintain the magnetic field structure without relying on the instability and hence improve the energy confinement performance.
- Magnetic mirror confinement *75
A kind of linear type device called open-end device in the magnetic confinement system. Currents flowing in the same direction along two circular coils produce a magnetic field distribution in which the field is strong beneath the coils and weak between the coils and the field lines are concentrated beneath the coils. This is called a magnetic mirror, since charged particles with a large velocity perpendicular to the field line are reflected where the magnetic field is strong and confined between the coils. On the other hand, charged particles with a large velocity parallel to the field line are lost at both ends of the magnetic mirror. This loss can be suppressed by setting additional mirrors at both ends and producing plasmas with a high electrostatic potential. This system, in which the plasma is confined with the magnetic field and the electrostatic potential, is called a tandem mirror device.
- GAMMA-10 *76
A complex mirror device operated at Plasma Research Center, University of Tsukuba, Japan. The length of the center mirror is 6 m, while the total length is about 27 m. In order to improve the confinement along axis, electron cyclotron resonance heating is applied to produce the ion confinement potential in the plug parts at both ends of the device.
- Potential formation *77
Since the potential formed in a plasma is related to confinement performance of the plasma, researches are carried out toward elucidating the mechanisms of the potential formation and confinement improvement by the potential.
- Compact torus *78
A system, which, unlike tokamak, does not use toroidal coils for the purpose of making device simple and compact, but confines plasma with torus-shaped closed magnetic field that is formed only after the current is conducted in the plasma. It includes reversed field mirror, field reversed configuration and spheromak.
- Magnetic reconnection *79
Magnetic reconnection is a change in the connection of magnetic field lines, which have mutually antiparallel components, due to the electrical resistivity etc. of plasma, and is a principal phenomenon that induces various nonlinear time evolution/relaxation phenomena such as self-organization in laboratory and astrophysical plasmas. Once the reconnection of magnetic fields occurs, plasmas on different field lines mix and the magnetic energy is converted into heat and kinetic energy of the plasma.
- Spherical torus *80
A machine in which the plasma aspect ratio A (= (plasma major radius) / (plasma minor radius)) is extremely small (A < 2), compared to that in usual tokamaks (A3). The beta values of several tens percent have been achieved experimentally, and its research contributes to the performance enhancement of the tokamak.
- TST-2 (Tokyo Spherical Tokamak - 2) *81
A spherical tokamak device at University of Tokyo. Bt = 0.2-0.4 T, Ip = 200 kA, aspect ratio A = 1.6 (>1.5).
- LATE (Low Aspect ratio Torus Experiment) *82
LATE (Low Aspect ratio Torus Experiment) is a low-aspect-ratio spherical tokamak device operated at Kyoto University (A = 1.3, Ip = 3 kA, elongation = 1.3). It is a spherical tokamak device without Ohmic heating but using only microwave in the GHz band. In particular, they mainly conduct unique research on generation and sustainment of a low-aspect-ratio spherical tokamak plasma without a center solenoid coil, using RF of electron cyclotron range of frequency.
- TS-3/TS-4 *83
A unique fusion plasma experiment device at University of Tokyo, in which generation and comparison of all current-carrying spherical torus systems including field reversed configuration, reversed field pinch, spheromak, and spherical tokamak are possible in a single device. As a significant feature, multiple tori as described above can be generated together to study their coalescence along the major axis and the magnetic reconnection.
- Internal conductor machine *84
This refers to a device which confines plasma by flowing current in the conductor located inside the plasma to make a magnetic field configuration for stable plasma confinement. Though it has engineering difficulties in application to a nuclear fusion reactor, it is expected to play an important role in basic research on nuclear fusion plasma such as magnetohydrodynamics and plasma confinement physics.
- Relaxation theory *85
Relaxation in plasma refers to a process where the plasma reaches a state in which the energy of the magnetic field is minimized through the state of macroscopic turbulence. This was confirmed in the reversed field pinch plasma.
- Proto-RT/Mini-RT *86
Proto-RT and Mini-RT are small internal conductor machines at University of Tokyo. Proto-RT with normal conduction coils is in operation, and Mini-RT with superconducting coils is under construction. In the Mini-RT, high-temperature superconductor strands are used for the magnetic levitation coil.
- CSTN-IV *87
A tokamak-type small plasma confinement device at Nagoya University. It can produce a plasma with a plasma current of 1 kA and an electron temperature of about 10-20 eV. It has characteristics, unique in the world, of high-repetition-rate discharges, and is capable of perfect alternating discharges and high-repetition-rate discharges with an operation duty of 50%, which enables simulation of a long duration discharge.
- NAGDIS-II (Nagoya University Divertor Simulation - II) *88
A linear divertor plasma simulation device at Nagoya University (Nagoya University Divertor Simulation). It was constructed to produce a high-heat-flux plasma of a long column with a large diameter. The steady high-heat-flux plasma allows studies on plasma-wall interaction in the divertor region.
- Detachment phenomenon *89
A state where the ion flux to a target plate disappears and the plasma heat flux decreases greatly. It is observed when the particle recycling is enhanced with gas puffing and then a high-density low-temperature plasma is formed in the divertor region. It is caused by recombination of the plasma particles and the momentum loss by the charge exchange reaction. In the detached state, the region where the neutral particles are ionized moves away from the vicinity of target plates. At the same time, the region with large radiation loss and charge exchange loss moves toward the main plasma and stagnates. It is proposed to utilize this phenomenon in order to reduce the heat flux to the target plates in a nuclear fusion reactor.
2.4 Fusion Technology Research
- Superconducting coils *90
Electrical magnets (coils) using superconductors. Since superconductors have zero electrical resistivity, unlike normal conductors, operation of superconducting magnets does not yield joule loss. Also, since superconductors can conduct large current even with small cross section, they are suitable for making a compact magnet.
- Niobium-tin *91
A kind of wire rod materials for superconducting coils. Its chemical formula is Nb3Sn. When it is cooled down to about -260°C, it enters the superconducting state where the electric resistance becomes zero. Its superconducting characteristic is very excellent compared with the niobium titanium. But since it is intermetallic compound, in order to produce the compound, it is necessary to heat-treat the conductor. It is very fragile as material, but owing to the recent progress in fabrication technology, it now comes to be used in coils for high magnetic field. It is adopted for the toroidal magnetic field coil and the central solenoid of ITER.
- Remote maintenance *92
Since the internal structures of vacuum vessel (blanket, divertor, etc.) are highly radioactivated by neutrons produced in fusion reaction, they are not directly accessible to human. Therefore, in case these structures are damaged, we have to maintain and repair them by remote manipulation using robot technology etc. Remote maintenance refers to these technologies and tools.
- Blanket *93
In a fusion reaction generator, the structure placed surrounding plasma in which nuclear fusion occurs is called blanket. Blanket bears either of the following functions such as: shielding of vacuum vessel and superconducting coils outside of it from neutrons; transmuting lithium compound to tritium by utilizing neutrons produced in fusion reaction; capturing neutrons and converting their energy to heat.
- In-vessel self-propelled maintenance system *94
This is one of the remote maintenance equipments. It is a system to maintain and repair the structural components in the vacuum vessel by supporting rails, laid in vacuum vessel at the maintenance time, from multiple points around the vacuum vessel and using self-propelled robots (vehicle type manipulators) etc. that can move along these supported rails.
- Bonding technique *95
Here, we mainly refer to the technology to bond the mutually different kind materials. For plasma facing components, bonding methods like brazing or diffusion bonding are used.
- Divertor plate *96
A device to guide charged particles flow out from main plasma to neutralizer plates (divertor plates) or to exhaust region (pump) so that they do not hit directly on the neighboring walls, by devising the shape of magnetic field lines around plasma. It is particularly effective to reduce impurities in plasma.
- Tritium Process Laboratory (TPL) *97
Tritium Process Laboratory at Japan Atomic Energy Research Institute. A facility for research and development on tritium engineering technology for nuclear fusion reactor development. It is the only facility in Japan (at present, the largest scale in the world in nuclear fusion research) to handle gram-level of large amount tritium (licensed amount of tritium storage is 22.2 PBq = about 63 g). Up to now, it has collected basic characteristics of major processes in the fuel system for fusion reactor (fuel purification, collection, isotope separation and storage), reflected these to the design of future fusion devices such as ITER, and accumulated the results of tritium safe handling and operation management of the facility for more than 15 years. Recently, it puts emphasis on the development of blanket-bred-tritium recovery system, and it is also widely utilized for the research on the prevention of tritium contamination and decontamination.
- System integration technology *98
During the operation of reactor, the components of fusion reactor are required to have safety and reliability under the combined environment of heat, particle, radiation, and electromagnetic force, that is not experienced in individual elemental technological tests. System integration technology refers to the technology to make these components function as an integrated system.
- Fusion Neutronics Source (FNS) *99
FNS (Fusion Neutronics Source) is an accelerator-type fusion neutron source located at Japan Atomic Energy Research Institute (Tokai Research Establishment). It accelerates deuterium ions to about 400 kV by an accelerator, injects them into a target adsorbed with tritium (T), and produces 14 MeV neutrons by DT reaction. The maximum beam current is 20 mA, maximum neutron production rate is 2 x 1012 n/s, and now its intensity is the highest in the world. Until now, measurement experiments of physical quantities which originate from the reaction of DT neutrons such as tritium generation rate, nuclear heating, induced radioactivity, shielding performance, and experimental research for the evaluation of calculation accuracy of nuclear design code, etc. have been carried out. At present, it is also actively utilized in researches such as irradiation effect of functional materials, experiment on the nuclear phenomenon of material damage, and development of the detector for nuclear instrumentation.
- Lithium compounds *100
Compounds of lithium. In a fusion reactor, they are used as materials to breed tritium (tritium breeder) in the reactor itself, which exists only in a small amount in nature. The candidate materials are lithium oxide (Li2O), lithium titanate (Li2TiO3), lithium silicate (Li4SiO4), lithium alloy (Li17Pb87), molten salt (FLiBe) and so on.
- Solid breeding blanket with packed pebble bed *101
Solid breeding blanket that utilizes small pebbles of tritium breeder and neutron multiplier. To maintain the heat and mechanical characteristics of the tritium breeder and neutron multiplier against the neutron irradiation effect, breeder and multiplier are packed as small pebbles.
- Irradiation damage (dpa) *102
Acronym of displacement per atom, and an index for the number of constituent atoms of the materials that are displaced from lattice points due to neutron irradiation. Thus, irradiation damage of 1 dpa implies that constituent atom of the materials is displaced once from the lattice point on average (this is really a matter of average, so individually, some constituent atoms may be displaced twice, while some others may not).
- Radioactivation *103
The materials of fusion devices (especially in-vessel components and structural materials) are radioactivated by interaction with fusion-produced neutrons. Their radioactivation level is determined comprehensively from the kind of radioactivation-produced nuclide (half-life, radiation species, energy, etc.) and their density. For example, in the disposal of radioactive wastes, depending on the radioactivation level (density level of radioactive nuclide), they are classified to high βγ waste, low level waste, very low-level waste and clearance-level, etc., and undergone appropriate disposal for each level.
- Material irradiation test facility *104
A test facility to examine the damage of materials and the resultant change in their characteristics due to neutron irradiation. At present, what is available for this purpose is mainly a nuclear fission reactor having ports for the irradiation tests. Since the material characteristic is sensitive to temperature, the control and measurement of temperature are important functions in the irradiation facility.
- Neutron *105
A neutral elementary particle. It composes an atomic nucleus with proton. The charge is 0, and the mass is 1.6749 x10-27 kg. Neutron alone is unstable and collapses to proton by β- decay with half-life of 12.5 minutes. Since neutron is electrically neutral and can easily enter into a nucleus, it is used to initiate nuclear reaction. The number of neutrons passing through unit area per second is called neutron flux and its unit is neutrons/m2/sec. Incidentally, the number of neutron particles per unit volume is called neutron number density or neutron density.
- International Energy Agency (IEA) *106
An international energy agency under OECD established in the wake of oil resource supply-demand problem etc. It is established in November 1974 with an aim to improve the supply structures. One of its main duties is to promote cooperation among the participating countries to develop alternative energy resources.
- International Fusion Materials Irradiation Facility (IFMIF) *107
An international fusion materials irradiation facility being designed in cooperation among Japan, US, EU, and Russia under IEA. When deuterium ions accelerated to 40 MeV irradiate a lithium target, neutrons with energy around 14 MeV are produced effectively as a result of nuclear reaction. By irradiating these neutrons to materials, tests on the change in their characteristics under the condition simulating the fusion reactor environment are conducted.
- Niobium aluminum wire rod material *108
One of the materials for superconducting coil wire rod recently in development and its chemical formula is Nb3Al. When cooled down to about -260°C, its electric resistance becomes zero and it enters the superconducting state. Its superconducting property is better than that of niobium tin. In particular, since the degradation by the mechanical strain imposed on the material is small, fabrication of coils can be simplified by adopting the react-and-wind method, in which wire winding is carried out after the production and heat-treat of niobium aluminum. At Japan Atomic Energy Research Institute, fabrication technique of niobium aluminum wire rod material has been developed using jelly-roll method.
- Energy conversion efficiency *109
Conversion efficiency from electrical energy to laser energy or from laser energy to nuclear fusion energy. In order that a laser fusion reactor be viable, the product of these two efficiencies needs to be more than ten.
- Laser-diode-pumped solid-state laser (DPSSL) *110
Laser-diode-pumped solid-state laser (DPSSL) uses laser diode (LD) as an excitation (pumping) source. Since it uses the luminescence of a narrow line spectrum that matches the absorption spectrum of those to be excited, it is efficient and can be repeatedly operated because unnecessary heat generation is small. In general, LD-pumped solid-state laser has achieved efficiency of more than 10% and high repeating operation of more than 10 Hz. From the later half of the 1980s, LD-pumped solid-state laser has rapidly advanced in output level and operational performance owing to the emergence of high-power high-efficiency LD. In continuous mode, laser output of more than 10 kW can be obtained.
- HALNA laser *111
HALNA (High Average power Laser for Nuclear fusion Application) is a laser system, which uses DPSSL now being developed, as a laser driver for fusion reactor. It aims at relatively high output energy at a repetition rate of about 10 Hz.
- Excimer (KrF) laser *112
A gas laser using rare gas halide, which operates due to the transition of grand-state molecule that has intrinsic antibonding nature and oscillates in ultraviolet (248 nm etc.). It utilizes the luminous phenomenon caused by the collapse of excited dimer. It is mainly used for semiconductor manufacturing (lithography).
- Irradiation effect *113
The change in materials caused by the irradiation of particle beam etc. The irradiation effect with 14 MeV neutron becomes a big problem in the fusion reactor. The irradiation by high-energy neutrons causes direct or indirect effects on materials, such as displacement damage, nuclear transmutation damage, and electronic excitation, etc. and by these the characteristics of materials change.
- Material irradiation damage modeling *114
Modeling of irradiation effects on materials. Changes in materials caused by irradiation have the following characteristics. The irradiation effects on materials start from the generation of point defect/ heterogeneous atom due to displacement damage or nuclear transmutation damage. The process of their annihilation by diffusion or the formulation of defect cluster by their coalescence causes microscopic structural change. Furthermore, it affects, for example, the motion of dislocation, which is a microscopic process of displacement, and changes strength characteristic, which is a macroscopic characteristic.
- Electromagnetic and thermal structural analysis *115
The vacuum vessel and the in-vessel structure of the fusion reactor are subjected to electromagnetic force (electromagnetic stress) and thermal stress corresponding to nuclear heating, specific to the nuclear fusion. It is necessary to design vacuum vessel and in-vessel structure so that they have sufficient structural soundness for these stresses. This term refers to the generic name of evaluation and analysis method for this design, such as on electromagnetic force, nuclear heating, and stress.
- Neutron transport *116
The spatial movement of neutrons produced in nuclear fusion, nuclear fission or other nuclear reaction, accompanied by the change in energy.
2.5 Fusion Reactor System Design
- Fusion reactor SSTR *117
Acronym for Steady State Tokamak Reactor (SSTR) in which 75% of the plasma current is driven by spontaneous current (bootstrap current). It is a reactor concept that enables steady-state operation by using spontaneous current. It was proposed by Japan Atomic Energy Research Institute in 1990 and became the prototype of the present steady-state tokamak reactor.
- Highly economically efficient reactor CREST *118
Acronym for the reversed shear configuration tokamak reactor, Compact Reversed Shear Tokamak, which aims at compactness and low cost. It is a reactor concept aiming at high economic efficiency, proposed by Central Research Institute of Electric Power Industry in 1997.
2.6 Safety Research
2.7 Achievements in Academic Fields
- Transition phenomenon *119
A phenomenon in which a system with specific state jumps suddenly to another state.
- Limit cycle *120
An attractor with finite length. This refers to the repetition of transition and back transition, and corresponds to a periodic motion.
- Laser nuclear physics *121
A discipline studying the nature of nuclei by affecting nuclei with intense electric field of ultra intense laser.
- High energy density physics *122
A science that explores the physics under extreme state of very high pressure obtained only at the center of star.
- First principle *123
Among the laws that rule physical phenomena, this refers to a more fundamental principle or law. In plasma physics, the equation of motion for charged particles and the law of Maxwell's electromagnetic field correspond to this.
- Magnetohydrodynamic instability *124
An instability caused by the collective motion of many charged particles composing plasma. In some case, it may bring the loss of plasma from the plasma confinement region.
- Electrostatic fluctuation *125
Generic term for the fluctuation that is excited even if the fluctuation of magnetic field is not accompanied. This includes the drift wave, the ion temperature gradient instability, the electron temperature gradient instability, and so on.
- Gyro-kinetic theory *126
A charged particle makes a movement twisting around a magnetic field line. This refers to a theoretical model, which describes the motion of plasma particle by focusing on its spiral motion (gyro motion). The equation of motion of charged particles is regarded as one of the first principles in plasma physics. However, it is impossible to predict the behavior of fusion plasma, which is a multi-particle system, directly by particle simulation. Therefore in this theoretical model, the freedom of multi-particle system is reduced with keeping the effect of the gyro motion that is important for the plasma behavior.
- MHD nonlinear dynamics *127
Magnetohydrodynamic behavior in which nonlinear effect becomes important. Recently it is considered important for explaining the sudden occurrence of instabilities or the growth rate of instabilities observed in experiments.
- Internal transport barrier *128
A phenomenon in which good plasma confinement region emerges locally within the plasma. In the vicinity of this barrier, density profile, temperature profile, etc. are observed to have steep gradients.
- Integrated simulation code for burning plasmas *129
A simulation code in which multiple physical simulation codes are combined and used, to resolve complicated plasma phenomena that are unpredictable by a single physical model or physical code, and gives quantitative prediction on fusion burning plasma.
2.8 Spin-off Effects on Industry
- Exhaust-gas-selection technology *130
In order to burn plasma stably for a long time, it is necessary to remove helium, the ash of nuclear fusion reaction, continuously. In the selection exhaust technology, unreacting fuel and helium in the mixed exhaust gas are continuously separated to each component using the difference in the adsorption affinity of zeolite adsorbent to each of helium and hydrogen isotope fuel. The fuel constituent is injected into core plasma again, to improve the fuel utilization efficiency. To be more specific, it is done by a vacuum pumping system composed of separation column filled with zeolite adsorbent and mechanical pumps such as turbomolecular pump. If the combination of the absorbent and the mixed gas is properly chosen, this technology can be applied also to the separation of a general mixed gas, and is applied to the separation and the collection of perfluorocarbon (PFC) gas that is a kind of global greenhouse gas.
- Pellet technology *131
A technology of making pellets developed in the laser fusion research. Particularly, since the sphericity of the fuel container for central ignition made by emulsion method reaches 99.98%, there is a research that utilizes this high accuracy and applies it to the impact sensor.
- Extreme ultraviolet light source *132
Extreme ultraviolet light (EUV) has wavelength that is more than one order of magnitude shorter than that of excimer laser (wavelength: 200 nm). It is regarded as a promising light source for the lithography of the next generation semiconductor manufacturing and its development is considered urgent.
2.9 Training and Education
2.10 International Collaboration
3.1 Strategy for Early Realization of Fusion Energy
- Pulse operation by electromagnetic induction *133
At the time tokamak type experiment device was invented, it was thought that only an intermittent operation was possible for tokamak because tokamak generates plasma by the principle of the transformer (electromagnetic induction). Such an intermittent method of operation is called a pulse operation.
- Steady-state burning plasma *134
This refers to a state where fusion core plasma*11 can be maintained steadily.
- Plant efficiency *135
The efficiency when the thermal energy generated by nuclear fusion reaction is converted to other form of energy such as electricity.
- Tritium Breeding Ratio (TBR) *136
In a fusion reactor, the ratio of the amount of tritium converted in the breeding blanket to the amount of burning tritium is called Tritium Breeding Ratio (TBR).
- Neutron fluence *137
Neutron flux is a number of neutrons that pass through the unit area per second, and its integration over a certain period is the integrated neutron flux, called fluence. Its unit is the number of neutron/m2. Along with absorbed radiation dose (Gy) and displacement per atom (dpa), etc., it is commonly used to show the amount of neutron irradiation for material degradation.
- Heat flux (High heat flux components) *138
A quantity of heat divided by the heated area and shows the heating power per unit area. It is one of the indices representing the load to the plasma facing components.
- Burning plasma control *139
Here burning plasma refers to the plasma in which fusion reaction is occurring using D and T as fuel. Burning plasma control refers to the control technique necessary to maintain plasma performance steadily in the self-ignition condition or the parameter region near to it.
- Current diffusion time *140
The current in plasma is mainly generated by the flow of the particles (electrons or ions) in the direction of magnetic field lines. The particles make a movement gyrating around magnetic field lines. But if they transfer to the different magnetic field lines by the scattering due to collisions, the current in plasma also spreads (diffuses) in the direction perpendicular to the magnetic field lines. The current diffusion time is a typical time constant for this to occur.
- Low sputtering *141
The phenomenon where a fast particle collides with the solid and the composing atom is flipped out from the surface of the solid is called sputtering. Low sputtering refers to the character that has low rate of material sputtering. The low sputtering material is suitable for the protection material (armor) of the plasma facing components, because the amount of erosion by the sputtering is small.
- Aspect ratio *142
Characteristic parameters of a toroidal magnetic confinement system are the radius of the torus (major radius R) and the fatness of the torus (minor radius a), and their ratio is called aspect ratio (A=R/a). Typical tokamak devices have an aspect ratio of A3, while helical devices have an aspect ratio of A=57. Recently, a very small aspect ratio tokamak (A<2) has attracted a great deal of attention and is called spherical tokamak.
- Plasma shape *143
Plasma confinement and stability properties are strongly dependent on the cross-sectional shape (triangularity, ellipticity). Thus, optimization of cross-sectional shape and its control become important.
- Feedback control *144
Generally, the feedback control is a control method to return the output signal to the input side (feedback), and then the output is controlled and corrected. In a fusion experimental device, this refers to the appropriate control of plasma using the measured density, temperature, and plasma current. For example, in the feedback control of density, this corresponds to the maintenance of constant density by measuring the plasma density and increasing (decreasing) the fuel gas input when the plasma density decreases (increases).
- Self-sufficiency of fuel *145
The fusion reactor has a feature that it can continue the operation without supplying tritium from outside by converting the lithium compound and producing tritium, in the breeding blanket, and using it as a fuel. The ability to continue operation without supplying the fuel from outside like this is called self-sufficiency of fuel.
- Working Group on Fusion Research of the Special Committee on Basic Issues of the Subdivision on Science under Council for Science and Technology*146
A working group set up under the Special Committee on Basic Issues of the Subdivision on Science under Council for Science and Technology, which is a council of the Ministry of Education, Culture, Sports, Science and Technology. It was in action from July 2001 until January 2003. Intensive deliberations on the centralization and improvement of efficiency in the fusion research were carried out by the fusion researchers who represented their specific research fields, and it compiled the report "Future Direction of National Fusion Research" on January 8, 2003.
3.2 Meaning and Positioning of Academic Research on Fusion
- Complexity science *147
This refers not only to physics such as fluid mechanics but also to the gcomplexh phenomena observed in the proliferation of the living organism such as fungi or in the economy etc. It is known that even a deterministic dynamical system exhibits complex behavior, which is impossible to express analytically. The fluid mechanics is studied as its typical subject of research because of its large degree of freedom and the existence of dissipation process.
3.3 Education and Training, and Sustainable Development of Fusion Basic Technology
3.4 Promotion of International Collaboration
3.5 Balance among various Projects and Check and Review
4.1 Developmental Research in Tokamak
- ITER International Fusion Energy Organization *148
This is the implementing entity of ITER project that is an international joint project. It constructs and operates ITER to achieve the burning plasma and to conduct comprehensive tests of the engineering technology.
- Fusion forum *149
It was established in 2002 as an organization of independent participation such as researchers, specialists and engineers of universities, research institutes, and the industrial world, and eminent persons from various fields, to support the promotion of research and technological development for the achievement of the nuclear fusion energy. It puts immediate emphasis on the activity concerning the issues and plan to promote the research and development in relation to ITER project and the participation in it.
- In-pile test *150
A test aiming to investigate the characteristics and to demonstrate the functionality of the materials and equipments under the irradiation environment (in a fusion reactor, mainly neutron). A test where the test object is put in a fission reactor to simulate the environment in a fusion reactor and to examine its characteristics. The nuclear fission reactor such as a light-water nuclear reactor or a fast breeding reactor is used for conducting the test.
- Neutronics test *151
Neutronics means various developmental researches to investigate the behavior of neutrons necessary for the design of nuclear fusion reactor. Neutronics tests are the experiments for fusion neutronics (fusion neutron technology). More specifically, this refers to the research on measurement of various physical quantities which originate from the reaction of DT neutrons, experimental evaluation on the accuracy of the nuclear design code, irradiation effects on functional materials for nuclear fusion reactor, experiments to elucidate nuclear phenomena of material damage, development of the detector for nuclear instrumentation, and research on the application of nuclear fusion neutron, etc.
- Heavy irradiation data *152
At the nuclear fusion power plant level, it is said that the neutron wall loading of the blanket during its lifetime becomes about 10-15 MWa/m2, and in displacement per atom, this corresponds to 100-150 dpa. Here, dpa is a unit for expressing the displacement damage (1 dpa means each atom in the material has been displaced once, on average). In existing nuclear reactor, fuel clad of fast breeder reactor receives the damage of about 100 dpa.@@Since many of the irradiation effects have cumulative nature, and in a fusion reactor, the effect of nuclear transformation of helium etc. will be added, acquisition of irradiation data is indispensable for predicting the change of material characteristic with good accuracy, and reflecting it to the design and official licensing.
- D-Li stripping reaction *153
A kind of nuclear reaction occurs when deuteron (nucleus of deuterium) is accelerated to a high speed and collides with lithium atom. It is called so, because it is a reaction process where either proton or neutron, which constitute deuteron, will be ripped off (strip). As a result of this reaction, left neutron (or proton) runs concentrated in a narrow angle along the direction of accelerated deuteron. Therefore, this reaction is useful to produce a neutron source that has strong forward directivity. In International Fusion Materials Irradiation Facility (IFMIF) project, proximate environment of fusion neutron (14 MeV) is simulated by using this reaction.
- Tritium process *154
Processing of tritium, which takes a variety of chemical forms such as molecule, water or methane, and is mixed with various materials. In the fuel circulation system of the fusion reactor, tritium process is carried out where non-burning tritium in the plasma exhaust gas is collected, purified, and reused as fuel.
- Tritium safety management technology *155
It is a technology to manage safely the tritium as a radioactive material. Tritium is easy to diffuse or leak as a radioactive gas. Therefore the technology to remove and collect tritium upon confining it in the space surrounded by barrier is essential.
- Nonlinear open system plasma *156
Even the confined tokamak plasma is substantially an open system, because plasma exchanges information with the outside system through particles and energy, etc. Since plasma is a nonlinear medium, a nonlinear process becomes important in the exchange of such information, and it exhibits characteristics specific to open system such as the structural formation.
- Numerical EXperiment of Tokamak (NEXT) *157
Here the discharge process of tokamak plasma is numerically and virtually embodied on computers, using numerical simulation codes based on various physical models describing the physical characteristics and physical phenomena of tokamak plasma.
4.2 Academic Researches in Fusion
- Nonlinear and far non-equilibrium medium *158
The medium whose system is not in the state of thermal equilibrium is called a non-equilibrium medium, and the system very far from the state of thermal equilibrium is called far non-equilibrium medium. Following this definition, the fusion plasma is a nonlinear and far non-equilibrium medium. Not in the statistical physical sense, but in the magnetohydrodynamical sense, magnetically confined plasma is generally in a stable equilibrium state.
- Autonomous structural formation *159
The autonomy that appears in the dissipative system by nonlinear effects. It has a meaning of gspontaneous structural formationh.
4.3 Sharing of Fusion Research and Development
4.4 Promotion of Talented Persons and Massage to Society
4.5 Global Structure of Research and Development and the Road to Utilization
4.6 Check and Review Items and Transition Condition to the Next Phase