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Research activities on fusion engineering in Japan
Fusion Engineering and Design 49 – 50 (2000) 41 – 52 www.elsevier.com/locate/fusengdes
Research activities on fusion engineering in Japan Kenzo Miya * Nuclear Engineering Research Laboratory, Graduate School of Engineering, The Uni6ersity of Tokyo, 2 -22 Shirakata-Shirane, Tokai, Naka, Ibaraki 319 -11 -6, Japan
Abstract In this article the dual structure of fusion researches in Japan is introduced in light of universities, National Institute of Fusion Science (NIFS) and Japan Atomic Energy Research Institute (JAERI). Two topics of fusion materials research and fabrication of the helical coils of large helical devices (LHD) in NIFS are introduced as activities supported by Monbusho. As voluntary based researches an application of high Tc superconducting tapes to suppression of plasma vertical instability and a high heat removal system at a divertor are also introduced. Finally, focussing on the development of innovative technology for a Demo reactor, a new strategy of technical R&D is proposed. Systematization of dependent technical knowledge and construction of a virtual plant concept are key activities supporting creation of the new technology. © 2000 Published by Elsevier Science B.V. Keywords: Fusion engineering; National Institute of Fusion Science; Japan Atomic Energy Research Institute; Japan
1. Introduction Structure of research activities on fusion engineering in Japan is shown in Fig. 1. Organizational aspects of the structure are characterized as. 1. Dual structures identified by. Competitive authorities of the Japanese government responsible for funding the researches in Japan i.e. the Ministry of Education, Science, Sports and Culture (Monbusho for brevity) and Science and Technology Agency (STA). Universities and National Institute of Fusion Science (NIFS) under auspices of * Tel.: +81-3-58417421; fax: + 81-3-58418631. E-mail address: [email protected] (K. Miya).
Monbusho and Research Institutes, JAERI and NRIM, under auspices of STA. 2. Research activities at universities were individual based on as usual style but group based as particular case in the period up to now at least. This style might change due to the unification of Monbusho and STA in 2001. 3. Basic research has been promoted primarily at the universities while facilities-dependent plasma physics researches or reactor oriented technologies has been promoted primarily at Japan Atomic Energy Research Institute (JAERI) involving activities at national institutes. Personally the author would recognize that the path of long-term R&D for commercialization of fusion energy is defined as step by step progresses of fusion technologies, which are categorized into three steps like,
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1. identification of fusion technologies in the elementary style — the first step; 2. component technologies satisfying specification of a reactor like machine international thermonuclear experimental reactor (ITER) — the second step; and 3. advanced technologies fulfilling economical requirements for a commercial reactor — the third step. In the second step the first two of three major missions of fusion energy R&D i.e. (1) engineering feasibility and (2) sufficient capability of safety, of a fusion reactor, thirdly (3) economical viability of a commercial reactor are going to be demonstrated. ITER will be an excellent opportunity where the engineering feasibility and the safety capability will be proved in the very evident way. ITER technologies have been developed at JAERI in collaboration with the Japanese industries and remarkable progresses have been already achieved in important areas of fusion technology. The author is sure that the R&D effort for reactor oriented technologies required by ITER and fundamental researches like those developed
at the Japanese universities should be in fact interacted to each other in the concrete way as they are in reality coupled in the deep structure in the manner of fundamental technical data being provided by universities researchers. The research activities at universities should by nature concentrate on systematization of fusion technology. We understand that Fig. 1 shows that the technological identification of a fusion reactor was realized through basic researches by the substantial contribution from the universities activities while the ITER technology was developed by the JAERI’s effort with the necessity of the bigger budget and the larger man-power to conduct the project. However, in view of future of fusion R&D, it is evident that we will encounter a most serious technical difficulties as to whether or not cost issues can be solved to reduce it to some reasonable level. This concern suggests that direct extrapolation of ITER technologies to a Demo reactor will not satisfy the economical requirement while the technical feasibility and the safety capability will have been demonstrated through the ITER project and thus enhancement of the economical capability should be achieved in a Demo reactor. If this would be possible, earlier commercialization of fusion energy shall be promising. Recognizing importance and difficulty of technologies for a Demo, it is anticipated very strongly that saving of human resources and R&D budget will be possible in the worldwide context of fusion community and as a result of this restructuring that sharing of necessary amount of tasks should be agreed and planned expecting significant contribution from not only larger institutes like JAERI but also universities. This is a new interest and should be favorably a new trend in the research style of fusion technology.
2. Recent progress in fusion engineering in Japanese universities
2.1. Material science and engineering Fig. 1. (a) Dual Structure of Japanese fusion reactor. (b) Fusion research activities at universities.
Research activities at Japanese universities have been very unique compared with US, EU and RF,
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focusing on both fundamental materials science and blanket engineering. There are two major research groups in Japan. One consists of university researchers and the other is a group of JAERI. Professors at seven major universities and NIFS have formed a kind of material research group and have been extending activities through network. They have meetings frequently in a form of seminar several times a year in addition to many occasions of informal meetings and discussions on networks. They participate in the activities developed in the framework of Japan/US collaboration called Japan US Program for Irradiation Test for Fusion Research (JUPITER) preceded by the former collaborative research program of FFTF/MOTA. Results from the fusion reactor materials research have been and are going to be published in many open publications. The major efforts can be seen in the publications. On-going irradiation experiments on structural and functional materials using domestic fission reactors, JOYO and JMTR, will be continuing and their post irradiation experiments and implementation of irradiation plans will also be continued. The successful accomplishment of in situ reactor studies will be followed by additional in-reactor experiments on resistively and optical property in JMTR. Post irradiation examinations of the varying temperature irradiation experiments will be carried out as a complementary to the JUPITER experiment. These experiments include many improvements of irradiation vehicle technologies. Simulation techniques of irradiation using ion accelerators will be improved by installations of new facilities.
2.1.1. Japan/US JUPITER project Post irradiation examinations (PIE) for the ATR-A1 experiment, which focused on low activation structural materials on both vanadium alloys and ferritic steels, are in progress. The ‘varying temperature experiment’ in HFIR has been successfully completed. ‘In situ thermal conductivity measurements’ in HFIR also started following experiment design, capsule design and fabrication. HFIR-RB experiment, at 300 and 500°C up to 10 dpa was completed and the cap-
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sules were disassembled and specimen shipment among PNL/ANL/IMR-Oarai has been performed. Assignments of Japanese scientists at PNL/ANL/ORNL and the US scientists at Tohoku University have been performed almost as planned. The irradiation for ‘varying temperature experiment’ and ‘in situ thermal conductivity measurements’ in HFIR will be completed soon. Post irradiation experiment of HFIR-RB irradiation will be the major task in the JUPITER project. Also, preparations for advanced irradiation experiments including ‘high temperature unshielded irradiation’ are underway.
2.1.2. De6elopment of low acti6ation materials [1,2] Materials development activities for irradiation resistant and reduced activation materials are extensively in progress. A low activation ferritics, JLF-1, has been characterized, as unirradiated and irradiated, by many university researchers as well as the information exchange with JAERI researchers who are focusing F82H low activation ferritic materials. Japanese universities researchers are leading vanadium alloy development and their database production under irradiation. Researches for fabrication of a large ingot of V–4Cr–4Ti alloy also started at NIFS. 2.1.3. Materials in reactor design Independent activities have been continued aiming to apply results of materials research to the design of ITER and other design activities, such as force free helical reactor (FFHR), steady state tokamak reactor (SSTR), drastically easy maintenance reactor (DREAM) and JT-60SU. Meetings on materials and fusion reactor design, such as US/Japan workshop on ‘power reactor design study’, were held and proceedings will be published. 2.1.4. R&D for intense neutron source de6elopment Japanese universities will start to plan R&D for intense neutron source development through the collaboration and coordination with JAERI
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2.1.5. IEA joint workshops To contribute to IEA joint workshops on fusion materials, such as ‘ferritic steels’, ‘vanadium alloys’, ‘SiC/SiC composites’, ‘refractory metals for fusion’, ‘theory and modeling’ and other related international meetings will be very important part of the annual program to enhance the efficiency of the domestic activities on fusion materials, especially, an international round-robin test of reduced activation ferritic steels, vanadium alloys and SiC/ SiC will become an important part of the international collaborative activities. 2.1.6. Modeling acti6ities Modeling activities are very unique features of Japanese University research on fusion materials. Fundamental understandings of mechanisms of material performance under fusion environments mainly from microscopic viewpoints are integrated to larger space and longer time scale models. Ultimate goal of the modeling is to predict material performance in fusion reactors. Required simulation technology such as molecular dynamics and Monte-Carlo simulations should be combined and verified by the carefully designed critical experiments. Irradiation studies using domestic irradiation facilities have been in progress. Irradiation with controlled temperature and irradiation histories were performed using JMTR. Irradiation studies with JOYO were also progressed, where a new temperature controlled irradiation rigs for JOYO have been made and specimen preparation is in progress. In situ resistivity measurement in JMTR was successfully performed. Mechanistic studies have been carried out using ion accelerators at the HIT facility at the University of Tokyo, Tokai site and other facilities, such as a tandetron accelerator facility of Research Institute for Applied Mechanics, Kyushu University and a HVEM-ion accelerator facility of Hokkaido University. University researchers are also using triple beam irradiation facility in JAERI, Takasaki to investigate transmutation effects in irradiated structural and functional materials. 2.2. Magnet technology A highlight of magnet technologies is placed on the construction and the successful operation of
the helical coils at NIFS [3]. This world largest superconducting device was conceived very difficult to construct with a very high precision of fabrication but the successful operation was achieved without any serious technical trouble. Characteristic features of the helical coils for large helical devices (LHD) at NIFS are summarized as follows: 1. Superconducting helical coils were successfully fabricated within a very high precision ( 92 mm) despite of very complicated geometry. 2. Malfunction of the coil such as helium gas leak, insufficient electrical insulation etc. did not take place up to now. 3. The very heavy structure of the coils, 850 tons could be cooled down uniformly resulting in an excellent coincidence with the design prediction. However, it should be mentioned that the specified design value of the coil current (13.0 kA) was not reached due to the heat generation in the coil in the transient so that the existing coil current is within 11.4 kA at this moment.
2.2.1. De6elopment of inno6ati6e component technologies The fabrication of such helical coils was first innovative technologies but also quality control in the whole manufacturing procedure was epoch making. As an example, we can point out the winding technology with use of a winding machine with computer 13 axis control. The winding work with the new machine should be combined with particular arrangement of working staffs possessing aspiring mentality and well-organized skills. Innovations made in the R&D of the helical coils are as follows item by item: 1. elucidation and evaluation of the magnetic resistance due to the Hall effect in the main stabilizer of pure aluminum; 2. development of the winding machine with computer 13 axes control to wind helical coil conductors of 36 km on site; 3. development GFRP of enhanced elasticity and low coefficient of thermal expansion (30 GPa and 0.3% for RT to 77 K);
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the nominal current is 13.0 kA (4.4 K/phase I) and 17.3 kA (1.8 K/phase II). The coil system is connected to the power system with high reliability, fault probability of less than 10 − 5, and coil quench protection system of time constant of 20 s. Total length of the conductor reaches to 36 km and it took 1 year and 6 months to complete the on-site windings. The mechanical accuracy of the helical coils is 92 mm for m/n= 1/1 deformation, which should be satisfied to keep from the formation of magnetic island and friction heat in pool-boiling-type SC conductor. Considerable accuracy level of the coil system was demonstrated by the electron beam mapping measurement with fluorescent mesh screen, which shows the expected magnetic structure of plasma. Fig. 2. Configuration of birds eye of large helical device.
2.3. New technology 4. technology development for control and prediction of deformation by electron beam welding of thick structures; 5. addition of tensile force during transverse shift between layers after winding; 6. precision treatment of ground insulation; and 7. advanced technologies for measurement of cable position and numerical control (NC) machining with utilization of laser. These innovative technologies result in the success of plasma operation as well as structural integrity of helical coils.
2.2.2. Technical o6er6iew of coils Bird’s eye view is shown in Fig. 2, where several key components such as a vacuum vessel, helical coils and poloidal coils are housed in the cryostat. We can understand that a structure of the helical coils is extremely complicated. The total weight is about 1500 ton and the LHe cooled mass is 850 ton. The required capability for the helium refrigerator is 2700 l/h at 4.4 K. The mass of the supporting structure for the LHD superconducting magnet system is more than 400 ton, which sustains the total magnetic force of 40 000 ton. In Table 1, specifications of the coils are shown. The conductor size is 12.5× 18.0 mm and
Some innovative technologies and concepts are proposed and investigated, which gives the vision for attractive fusion reactors. Among them, the author would like to pick up two concepts related to small-sized fusion reactors and high heat flux handling.
Table 1 Superconducting coils of LHD Items
Helical coil
Poloidal coil (IV/IS/OV)
Superconductor Conductor type
NbTi/Cu Compacted strands 3.90 120
– Cable-in-conduit
Major radius (m) Weight per coil (ton) Maximum field in 6.9 coil (T) Stored energy (GJ) 0.92 Nominal current 13.0 (kA) Conductor length 36 (km) Cooling method Pool-cooled Coil temperature 4.4 (K)
1.80/2.82/5.55 16/25/45 6.5/5.4/5.0 0.16/0.22/0.61 20.8/21.6/31.3 5.4/7.4/10 Forced flow 4.5–4.8
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K. Miya / Fusion Engineering and Design 49–50 (2000) 41–52 Table 2 Major parameters of HTSC tokamak HTSC tokamak Major radius (m) Minor radius (m) Elongation Plasma current (MA) Toroidal field at plasma center (T) Fusion power (MW) Net energy gain
Fig. 3. Structural cross section of HTSC tokamak and layout of HTSC coils.
2.3.1. High temperature superconducting (HTSC) plasma stabilizer [4] One of the concepts is HTSC plasma stabilizer, which passively improve the plasma instability. The electromagnetic force due to type-II superconductors, to which high temperature superconductors belong, is dynamically stable because of buildup of induced current in superconductors. The stability of tokamak plasmas is attained by the induced current suppressing motion of a current carrying plasma column, as is shown in Fig. 3. The induced current is generated to keep constant the magnetic flux penetrating the HTSC coils. Advantages of the stabilizing system include the followings compared with conventional ones. It requires no power supply unlike feedback control systems. Its stabilizing effect on plasmas is free of decay, which is inevitable in that of eddy current induced in structures. Considerable plasma stabilization owing to HTSC coils provides the wider range of plasma elongation, which is usually restricted to moderate
6.0 1.88 2.1 20 6.4 650
ITER/RCO 6.2 1.9 1.67 13.3 5.51 500 10
values because of the vertical instability. In other words, HTSC coils enable compact reactors with highly elongated plasmas. Some attractive designs with use of HTSC coils were made, which displayed the feasibility of the HTSC coils for plasma stabilization and its potential to enable highly elongated plasma. This is an HTSC tokamak which is an ignition machine with the same major radius as ITER/RCO (R=6.1 m), and its cross-section and major design parameters are shown Table 2 and Fig. 3, respectively. Assumptions adopted in the design of HTSC tokamak are to use the same data base and formula as ITER in principle and not to use any innovative technology other than HTSC stabilizing coils. Employing 0-d plasma analysis, we determined major parameters such as the major radius, the aspect ratio, the elongation and the plasma current. The elongation of 2.1 is chosen to the maximum value in the range that plasma positional stability is secured with HTSC coils. Fig. 4 shows a plasma operation contour (POPCON) plot, which provides an indication of plasma operation point of Q= . To generate the contours of the auxiliary power required for plasma energy balance, the ITER-89P confinement scaling with a multiplier, H = 2.0 were employed. If we take a fusion power of 650 MW, a contour line of 650 MW gives n= 1020 m − 3 T= 14 keV as the cross point with the curve of Paux = 0. To confirm that the plasma positional stability is secured, plasma stability analysis was conducted with use of a plasma response code with
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radial and vertical directions, respectively, and the centroid moves back to the original point, followed by the recovery of bp and li. These results indicate that the plasma is stabilized by HTSC coils sufficiently, which supports for the enough plasma stabilization owing to HTSC coils and feasibility of the design of HTSC tokamak.
2.3.2. Pebble drop di6ertor [5] The pebble drop divertor is a new concept applicable to the environment of very high heat load ( \ 20 MW/m2) and particle loading (\1024 atoms per m2 s). In the pebble drop divertor system, a large number of small divertor pebbles Fig. 4. Decision of operation point in POPCON plot for HTSC tokamak.
an HTSC current analysis. Fig. 5 shows time evolution of plasma against the disturbance instantaneous bp drop of 0.20 and li drop of 0.08 recovering in 5 s. The maximum deviations of plasma centroid are −0.25 and − 0.11 m in
Fig. 5. Response of various plasma parameters during plasma disturbances.
Fig. 6. Pebble drop divertor concepts and components. (a) Multi-layer coating pebble. (b) Schematic drawing of pebble cycle.
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Fig. 7. The maximum heat flux and the surface temperature rise of divertor pebble in typical operation of the pebble drop divertor. (a) Maximum heat flux as a function of pebble radius (irradiation time, 30 ms). (b) Surface temperature rise during irradiation at incident heat flux of 30 MW/m2.
are dropped in the divertor space to form a curtain as shielding the striking zone of divertor. The pebbles are coated with at least two layers as shown in Fig. 6a, which satisfies the requirements for plasma facing material. A kernel is made from graphite or low-Z refractory ceramic like SiC, BeO or Al2O3. The mechanical strength, the thermal stress and the heat capacity of a pebble depend on the kernel. Thin plasma facing layer,
which results in low erosion character can be allowed, since the irradiation period of falling pebble is very short. Moreover, eroded pebbles can be removed at the outside the divertor. Therefore, conventional low-Z materials and coating techniques can be applied for plasma facing layer. The pebble drop divertor system consists of (1) a drop controller array, (2) pebble processing systems including a regeneration stage, a heat exchanger and the separation stage of eroded pebbles, (3) a conveying system, which is schematized in Fig. 6b. Outstanding features of the pebble drop divertor system with multi-layer pebbles and the external pebble processing systems includes, 1. pebble conveying without the MHD effect; 2. simultaneous removal of incident heat flux and particle flux; 3. both pumping function and low bulk tritium are realized by tritium permeation barrier; 4. the feasibility of divertor plasma control by mixing various pebbles with different PFL; and 5. continuous replacement of eroded divertor surface. The feasibility of the pebble drop divertor system should be verified in view of thermal and mechanical properties. Fig. 7a shows the relation between the maximum permissible heat flux on the surface of pebble and the radius of the kernel. The permissible heat flux is estimated such that the surface stress does not exceed the compressive strength of kernel material. Fig. 7b is the rise of surface temperature at the divertor heat flux of 30 MW/m2. The results imply that pebble drop divertor system with pebble of 0.5–1 mm in radius can handle the high heat flux.
3. Discussion on strategy of fusion technology development
3.1. Requirements of fusion energy commercialization Basically speaking, the effort for fusion engineering will be completed when three kinds of feasibilities, i.e. technology, safety and economics,
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are fully satisfied. In this regard, it would be reasonable to consider that technical feasibility and safety capability necessary for a future fusion reactor will be demonstrated by a success of the ITER project. This verification of the two feasibilities with a reactor like plant is really important in the long-term fusion R&D effort. Along this line, the R&D efforts have been made for the ITER technologies as the seven elements of technology development [6,7] and are close to completion. In this context, however, we are confronted with a very crucial issue regarding a viability of a fusion reactor, i.e. how we can solve the remaining issue of cost. It would therefore be right to recognize that in-depth effort of technology is required to conquer the very difficult technology for an economical commercial reactor. Purpose of the special effort for a Demo is dual. One is a possibility to change definition of the role of a Demo as demonstration of an economical reactor from the previous definition of a Demo reactor for a proof of safety capability. A remarkable benefit from the new definition is that commercialization of fusion energy gets faster by more than one decade due to omission of the effort for safety issue of a Demo. Schedules of ITER and a Demo are shown tentatively in Fig. 8. An important aspect in Fig. 8 is that a considerable technical R&D effort for a Demo will start together with ITER construction and the Demo will generate electricity continuously with high availability in 30 years after the construction. Another purpose is to re-organize structure of international collaboration, which could significantly save man power and budget leading to avoidance of double or triple investments of them to development of innovative technology for an
Fig. 8. R&D schedule for ITER and Demo reactor.
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economical fusion reactor. International task sharing should be favored as an optimum of efforts. If the ITER project will be initiated, the number of fusion researchers is supposed to decrease. To avoid this unfavorable situation, parallel paths to the commercialization of fusion energy, as is shown in Fig. 8, should be taken to realize the above two benefits so that the reduction of scale of fusion research can be avoided.
3.2. Strategy of fusion technology Next will arise an important and difficult question of how we can solve the technology issue enabling an economical reactor. In recognition of the technical difficulty of a fusion power reactor, the committee on fusion research of Science Council of Japan founded a subcommittee to deal with this issue to develop cost-effective component technology. The subcommittee discussed technical issues for cost effectiveness for 10 months since October 1998 and is finalizing a report. Essential part of the report seems important from a viewpoint of strategy for fusion technology development and is introduced in the following. In general, there is a firm belief in inherent capability of technology progress. According to the very old Japanese newspaper, it predicted 23 technical items for future technology as promising to come true 100 years later. Some of the predicted technologies realized within 100 years were a long-distance call, a jumbo jet, a shinkansen (Japanese fast train), TV, air-condition, etc. Twenty items of them were realized now to our surprise. Thus, we can believe that fusion technology necessary for a commercial reactor will be achieved in due time in view of supports from very vapid progress of technology in other fields. We can point out an example. As introduced previously, a progress of high temperature super conductor technology is very rapid and at present it is possible to fabricate a superconducting small magnet of several Tesla with HTSC tapes. If the technology of HTSC will be technically available for toroidal and poloidal coil systems in the near future, performance of a fusion reactor will be improved very much leading to a cost effective
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maintainability and operationability. Another example is of course development of excellent structural material. There is a structured order to construct a framework of the technology development for an enhanced fusion reactor. We had better speculate a strategic frame, consisting of systematization of fusion reactor technology and of a construction of a virtual fusion power plant. The virtual plant concept consists of a material virtual laboratory as a first step, next an information virtual laboratory as a second one and finally a creation virtual laboratory as an integration of them. They are shown in Fig. 9 and it is important to notice that the first level is enveloped in the second one and supports it. The first and second levels are included in the third level and utilized fully in a system integrating them.
Fig. 9. Strategic frame for systematization of fusion reactor technology.
reactor by achievement of higher magnetic field. In fact fusion power is proportional to toroidal magnetic field. Furthermore change of cryogenic coolant from liquid helium to liquid nitrogen makes a fusion system simpler leading to facile
3.2.1. Systematization of fusion technology There are several diversed notions about the definition of systematization of fusion technology so that the author would define it as follows. 1. As fundamental feature of the systematization, it takes the same pattern as the langue-parole form common in the linguistics and enables technological creation based on the systematized paradigm of fusion technology. Construction of the paradigm is a goal of our effort. 2. It is possible to predict physical fusion phenomena in materials and technical performance of a fusion system or a fusion subsystem if one consults with the systematized fusion technology paradigm. At this moment, technical R&D for ITER has been conducted for many elements of it. These were conducted to achieve targets set for the ITER technology within well-defined constraints or with boundary conditions. Unless systematization is tried, we will have to do similar things again in future when we are required to provide similar results for different technical conditions. This is a waste of time and resources and should be avoided. Consequently, technical efforts have to be made at the same time to construct theory of fusion engineering referring to experimental contents and results of the above technical R&D.
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3.2.2. De6elopment of integrated fusion technology There are three characteristics in the integrated fusion technology, i.e. (a) capability of progress, (b) structuralization and (c) integration. In Fig. 9 is shown structure of the integrated fusion technology. It consists of three levels. In the first level, various interactions between materials of components and general loadings (thermal, mechanical, particle and electromagnetic environmental ones) are of design concern and prediction capability of their behaviors should be secured to some extent of accuracy. The material –loading interactions are observed in most structural and functional components of a reactor. By constructing methods of the predictions we could consider that identification of fusion technology is established. Most of the researches conducted up to now are in this level. Still now, very important technical issues are not solved. Neutron irradiation problem of structural materials is a typical one of them. An important element in this level is to find scientific or technical rules and theory to establish ways of prediction of phenomena observed in a fusion reactor. In the second level, an information laboratory is constructed virtually, where resources of computer systems are utilized fully leading to saving up of R&D budget and speeding up of technical R&D significantly. In other words, simulation capability are applied to analysis of non-linear and complex phenomena observed in components of a real fusion machine and minimized experiments are conducted to verify validity of the simulation. If we can get sufficient agreement between two we judge that prediction capability of long-term phenomena like degradation of materials is secured and then we proceed toward a very important phase of ‘controlling phenomena’. Examples of controlling phenomena are to reduce tritium inventory and to improve strength of structural materials by some technique, to decrease heat load to a diverter plate by adding useful elements, etc. In the third level, integrated functions of a reactor are of concern. They are an assurance of long-term safety capability, remote maintenance capability, avoidance of troubles and accidents
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and finally achievement of economical requirements. When we consider how these integrated functions of a reactor are satisfied, we are led to an introduction of a virtual plant concept in the strategic frame of the systematized research plan. The virtual plant proposed here is different from a conventional concept in two points. One is an introduction of a real time in the virtual plant. This is possible by introduction of ‘degradation of materials’ used in a plant. And another is the possibility to visualize major fusion phenomena inside a reactor and situations of reactor components at any time using systematic simulation tools and other means. Coincidence between a real plant and a virtual one should be secured at necessary time intervals using measurement data set of the real plant, which is a necessary input to various simulation codes. Once such a virtual plant is constructed, we can conduct virtual experiments arbitrarily with the model by changing design parameters and observing reactor responses in expectation of new technical findings and other similar things. In order to introduce a real time in a reactor, we need two different informations on degradation behaviors of components with time and history of loading conditions and environment where components of concern are subjected. If the two informations are available fully, it might become possible to predict any accidents to some extent of accuracy. This is of course a very tough task but we are required to challenge this goal. Economical improvement is expected possibly by a simulation study with the virtual plant through forward and feed back manipulation where significance of a new technology is also evaluated as to contribution, cost reduction and high performance of a reactor. At the same time R&D requirements are manifested and will be carried out at a proper organization and by organized researchers.
4. Conclusion 1. Up to now administration and organization of fusion research in Japan have been conducted under auspices of Monbusho and STA.
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2. However, both will be unified in 2001 and restructuring of the fusion research system will be discussed and different shapes of it might appear, according to a progress of the situation. 3. Steady progress of fusion technology at university has been made aiming at the systematization of it for long-term effort. 4. For restructuring of fusion technology researches, the new concept of integrated fusion technology is defined and proposed based on the discussions at the committee on Fusion Research, Science Council of Japan through associated systematization of fusion technology. 5. Development of a fusion virtual plant is an excellent approach and must be different from a conventional concept of a virtual plant in introduction of real time and visualization of working state of the plant at any time. Real time can be realized through an introduction of degradation parameter of component and material. Acknowledgements The author would like to express his sincere appreciation to a contribution by Dr T. Uchimoto
.
to the paper. Also a useful discussion with Professor A. Kohyama and Professor N. Sekimura about fusion material should be noted here. Professor Imagawa and Professor Motojima should also be noted here for providing materials on the helical coils.
References [1] A. Kohyama, H. Hishimura, D.S. Elles, R.L. Klueh, W. Dietz, K. Ehrlich, J. Nucl. Mater. 233 – 237 (1996) 138. [2] H. Matsui, K. Fukumoto, D.L. Smith, H.M. Chung, W. van Witzenburg, J. Nucl. Mater. 92 (1996) 233 – 237. [3] O. Motojima, et al., Progress summary of LHD engineering design and construction, Fusion Energy 1998, IAEACN-69/FT2/1, IAEA, Vienna, 1999 (in press). [4] T. Uchimoto, K. Miya, Application of high temperature superconductors to enhance nuclear fusion reactors, Fusion Technol. 36 (1) (1999) 92 – 103. [5] M. Isobe, K. Matsuhiro, Y. Ohtsuka, Y. Ueda, M. Nishikawa, Conceptual design of pebble drop divertor, Fusion Energy 1998, IAEA-CN-69/FTP/26, IAEA, Vienna, 1999 (in press). [6] R. Jayakumar, et al., ITER central solenoid manufacturing R&D, Fusion Energy 1998, IAEA-CN-69/ITERP1/21, IAEA, Vienna, 1999 (in press). [7] M. Nakahira, et al., Integration test of ITER full-scale vacuum vessel sector, Fusion Energy 1998, IAEA-CN-69/ ITERP1/24, IAEA, Vienna, 1999 (in press).
Research activities on fusion engineering in Japan Kenzo Miya * Nuclear Engineering Research Laboratory, Graduate School of Engineering, The Uni6ersity of Tokyo, 2 -22 Shirakata-Shirane, Tokai, Naka, Ibaraki 319 -11 -6, Japan
Abstract In this article the dual structure of fusion researches in Japan is introduced in light of universities, National Institute of Fusion Science (NIFS) and Japan Atomic Energy Research Institute (JAERI). Two topics of fusion materials research and fabrication of the helical coils of large helical devices (LHD) in NIFS are introduced as activities supported by Monbusho. As voluntary based researches an application of high Tc superconducting tapes to suppression of plasma vertical instability and a high heat removal system at a divertor are also introduced. Finally, focussing on the development of innovative technology for a Demo reactor, a new strategy of technical R&D is proposed. Systematization of dependent technical knowledge and construction of a virtual plant concept are key activities supporting creation of the new technology. © 2000 Published by Elsevier Science B.V. Keywords: Fusion engineering; National Institute of Fusion Science; Japan Atomic Energy Research Institute; Japan
1. Introduction Structure of research activities on fusion engineering in Japan is shown in Fig. 1. Organizational aspects of the structure are characterized as. 1. Dual structures identified by. Competitive authorities of the Japanese government responsible for funding the researches in Japan i.e. the Ministry of Education, Science, Sports and Culture (Monbusho for brevity) and Science and Technology Agency (STA). Universities and National Institute of Fusion Science (NIFS) under auspices of * Tel.: +81-3-58417421; fax: + 81-3-58418631. E-mail address: [email protected] (K. Miya).
Monbusho and Research Institutes, JAERI and NRIM, under auspices of STA. 2. Research activities at universities were individual based on as usual style but group based as particular case in the period up to now at least. This style might change due to the unification of Monbusho and STA in 2001. 3. Basic research has been promoted primarily at the universities while facilities-dependent plasma physics researches or reactor oriented technologies has been promoted primarily at Japan Atomic Energy Research Institute (JAERI) involving activities at national institutes. Personally the author would recognize that the path of long-term R&D for commercialization of fusion energy is defined as step by step progresses of fusion technologies, which are categorized into three steps like,
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1. identification of fusion technologies in the elementary style — the first step; 2. component technologies satisfying specification of a reactor like machine international thermonuclear experimental reactor (ITER) — the second step; and 3. advanced technologies fulfilling economical requirements for a commercial reactor — the third step. In the second step the first two of three major missions of fusion energy R&D i.e. (1) engineering feasibility and (2) sufficient capability of safety, of a fusion reactor, thirdly (3) economical viability of a commercial reactor are going to be demonstrated. ITER will be an excellent opportunity where the engineering feasibility and the safety capability will be proved in the very evident way. ITER technologies have been developed at JAERI in collaboration with the Japanese industries and remarkable progresses have been already achieved in important areas of fusion technology. The author is sure that the R&D effort for reactor oriented technologies required by ITER and fundamental researches like those developed
at the Japanese universities should be in fact interacted to each other in the concrete way as they are in reality coupled in the deep structure in the manner of fundamental technical data being provided by universities researchers. The research activities at universities should by nature concentrate on systematization of fusion technology. We understand that Fig. 1 shows that the technological identification of a fusion reactor was realized through basic researches by the substantial contribution from the universities activities while the ITER technology was developed by the JAERI’s effort with the necessity of the bigger budget and the larger man-power to conduct the project. However, in view of future of fusion R&D, it is evident that we will encounter a most serious technical difficulties as to whether or not cost issues can be solved to reduce it to some reasonable level. This concern suggests that direct extrapolation of ITER technologies to a Demo reactor will not satisfy the economical requirement while the technical feasibility and the safety capability will have been demonstrated through the ITER project and thus enhancement of the economical capability should be achieved in a Demo reactor. If this would be possible, earlier commercialization of fusion energy shall be promising. Recognizing importance and difficulty of technologies for a Demo, it is anticipated very strongly that saving of human resources and R&D budget will be possible in the worldwide context of fusion community and as a result of this restructuring that sharing of necessary amount of tasks should be agreed and planned expecting significant contribution from not only larger institutes like JAERI but also universities. This is a new interest and should be favorably a new trend in the research style of fusion technology.
2. Recent progress in fusion engineering in Japanese universities
2.1. Material science and engineering Fig. 1. (a) Dual Structure of Japanese fusion reactor. (b) Fusion research activities at universities.
Research activities at Japanese universities have been very unique compared with US, EU and RF,
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focusing on both fundamental materials science and blanket engineering. There are two major research groups in Japan. One consists of university researchers and the other is a group of JAERI. Professors at seven major universities and NIFS have formed a kind of material research group and have been extending activities through network. They have meetings frequently in a form of seminar several times a year in addition to many occasions of informal meetings and discussions on networks. They participate in the activities developed in the framework of Japan/US collaboration called Japan US Program for Irradiation Test for Fusion Research (JUPITER) preceded by the former collaborative research program of FFTF/MOTA. Results from the fusion reactor materials research have been and are going to be published in many open publications. The major efforts can be seen in the publications. On-going irradiation experiments on structural and functional materials using domestic fission reactors, JOYO and JMTR, will be continuing and their post irradiation experiments and implementation of irradiation plans will also be continued. The successful accomplishment of in situ reactor studies will be followed by additional in-reactor experiments on resistively and optical property in JMTR. Post irradiation examinations of the varying temperature irradiation experiments will be carried out as a complementary to the JUPITER experiment. These experiments include many improvements of irradiation vehicle technologies. Simulation techniques of irradiation using ion accelerators will be improved by installations of new facilities.
2.1.1. Japan/US JUPITER project Post irradiation examinations (PIE) for the ATR-A1 experiment, which focused on low activation structural materials on both vanadium alloys and ferritic steels, are in progress. The ‘varying temperature experiment’ in HFIR has been successfully completed. ‘In situ thermal conductivity measurements’ in HFIR also started following experiment design, capsule design and fabrication. HFIR-RB experiment, at 300 and 500°C up to 10 dpa was completed and the cap-
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sules were disassembled and specimen shipment among PNL/ANL/IMR-Oarai has been performed. Assignments of Japanese scientists at PNL/ANL/ORNL and the US scientists at Tohoku University have been performed almost as planned. The irradiation for ‘varying temperature experiment’ and ‘in situ thermal conductivity measurements’ in HFIR will be completed soon. Post irradiation experiment of HFIR-RB irradiation will be the major task in the JUPITER project. Also, preparations for advanced irradiation experiments including ‘high temperature unshielded irradiation’ are underway.
2.1.2. De6elopment of low acti6ation materials [1,2] Materials development activities for irradiation resistant and reduced activation materials are extensively in progress. A low activation ferritics, JLF-1, has been characterized, as unirradiated and irradiated, by many university researchers as well as the information exchange with JAERI researchers who are focusing F82H low activation ferritic materials. Japanese universities researchers are leading vanadium alloy development and their database production under irradiation. Researches for fabrication of a large ingot of V–4Cr–4Ti alloy also started at NIFS. 2.1.3. Materials in reactor design Independent activities have been continued aiming to apply results of materials research to the design of ITER and other design activities, such as force free helical reactor (FFHR), steady state tokamak reactor (SSTR), drastically easy maintenance reactor (DREAM) and JT-60SU. Meetings on materials and fusion reactor design, such as US/Japan workshop on ‘power reactor design study’, were held and proceedings will be published. 2.1.4. R&D for intense neutron source de6elopment Japanese universities will start to plan R&D for intense neutron source development through the collaboration and coordination with JAERI
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2.1.5. IEA joint workshops To contribute to IEA joint workshops on fusion materials, such as ‘ferritic steels’, ‘vanadium alloys’, ‘SiC/SiC composites’, ‘refractory metals for fusion’, ‘theory and modeling’ and other related international meetings will be very important part of the annual program to enhance the efficiency of the domestic activities on fusion materials, especially, an international round-robin test of reduced activation ferritic steels, vanadium alloys and SiC/ SiC will become an important part of the international collaborative activities. 2.1.6. Modeling acti6ities Modeling activities are very unique features of Japanese University research on fusion materials. Fundamental understandings of mechanisms of material performance under fusion environments mainly from microscopic viewpoints are integrated to larger space and longer time scale models. Ultimate goal of the modeling is to predict material performance in fusion reactors. Required simulation technology such as molecular dynamics and Monte-Carlo simulations should be combined and verified by the carefully designed critical experiments. Irradiation studies using domestic irradiation facilities have been in progress. Irradiation with controlled temperature and irradiation histories were performed using JMTR. Irradiation studies with JOYO were also progressed, where a new temperature controlled irradiation rigs for JOYO have been made and specimen preparation is in progress. In situ resistivity measurement in JMTR was successfully performed. Mechanistic studies have been carried out using ion accelerators at the HIT facility at the University of Tokyo, Tokai site and other facilities, such as a tandetron accelerator facility of Research Institute for Applied Mechanics, Kyushu University and a HVEM-ion accelerator facility of Hokkaido University. University researchers are also using triple beam irradiation facility in JAERI, Takasaki to investigate transmutation effects in irradiated structural and functional materials. 2.2. Magnet technology A highlight of magnet technologies is placed on the construction and the successful operation of
the helical coils at NIFS [3]. This world largest superconducting device was conceived very difficult to construct with a very high precision of fabrication but the successful operation was achieved without any serious technical trouble. Characteristic features of the helical coils for large helical devices (LHD) at NIFS are summarized as follows: 1. Superconducting helical coils were successfully fabricated within a very high precision ( 92 mm) despite of very complicated geometry. 2. Malfunction of the coil such as helium gas leak, insufficient electrical insulation etc. did not take place up to now. 3. The very heavy structure of the coils, 850 tons could be cooled down uniformly resulting in an excellent coincidence with the design prediction. However, it should be mentioned that the specified design value of the coil current (13.0 kA) was not reached due to the heat generation in the coil in the transient so that the existing coil current is within 11.4 kA at this moment.
2.2.1. De6elopment of inno6ati6e component technologies The fabrication of such helical coils was first innovative technologies but also quality control in the whole manufacturing procedure was epoch making. As an example, we can point out the winding technology with use of a winding machine with computer 13 axis control. The winding work with the new machine should be combined with particular arrangement of working staffs possessing aspiring mentality and well-organized skills. Innovations made in the R&D of the helical coils are as follows item by item: 1. elucidation and evaluation of the magnetic resistance due to the Hall effect in the main stabilizer of pure aluminum; 2. development of the winding machine with computer 13 axes control to wind helical coil conductors of 36 km on site; 3. development GFRP of enhanced elasticity and low coefficient of thermal expansion (30 GPa and 0.3% for RT to 77 K);
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the nominal current is 13.0 kA (4.4 K/phase I) and 17.3 kA (1.8 K/phase II). The coil system is connected to the power system with high reliability, fault probability of less than 10 − 5, and coil quench protection system of time constant of 20 s. Total length of the conductor reaches to 36 km and it took 1 year and 6 months to complete the on-site windings. The mechanical accuracy of the helical coils is 92 mm for m/n= 1/1 deformation, which should be satisfied to keep from the formation of magnetic island and friction heat in pool-boiling-type SC conductor. Considerable accuracy level of the coil system was demonstrated by the electron beam mapping measurement with fluorescent mesh screen, which shows the expected magnetic structure of plasma. Fig. 2. Configuration of birds eye of large helical device.
2.3. New technology 4. technology development for control and prediction of deformation by electron beam welding of thick structures; 5. addition of tensile force during transverse shift between layers after winding; 6. precision treatment of ground insulation; and 7. advanced technologies for measurement of cable position and numerical control (NC) machining with utilization of laser. These innovative technologies result in the success of plasma operation as well as structural integrity of helical coils.
2.2.2. Technical o6er6iew of coils Bird’s eye view is shown in Fig. 2, where several key components such as a vacuum vessel, helical coils and poloidal coils are housed in the cryostat. We can understand that a structure of the helical coils is extremely complicated. The total weight is about 1500 ton and the LHe cooled mass is 850 ton. The required capability for the helium refrigerator is 2700 l/h at 4.4 K. The mass of the supporting structure for the LHD superconducting magnet system is more than 400 ton, which sustains the total magnetic force of 40 000 ton. In Table 1, specifications of the coils are shown. The conductor size is 12.5× 18.0 mm and
Some innovative technologies and concepts are proposed and investigated, which gives the vision for attractive fusion reactors. Among them, the author would like to pick up two concepts related to small-sized fusion reactors and high heat flux handling.
Table 1 Superconducting coils of LHD Items
Helical coil
Poloidal coil (IV/IS/OV)
Superconductor Conductor type
NbTi/Cu Compacted strands 3.90 120
– Cable-in-conduit
Major radius (m) Weight per coil (ton) Maximum field in 6.9 coil (T) Stored energy (GJ) 0.92 Nominal current 13.0 (kA) Conductor length 36 (km) Cooling method Pool-cooled Coil temperature 4.4 (K)
1.80/2.82/5.55 16/25/45 6.5/5.4/5.0 0.16/0.22/0.61 20.8/21.6/31.3 5.4/7.4/10 Forced flow 4.5–4.8
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K. Miya / Fusion Engineering and Design 49–50 (2000) 41–52 Table 2 Major parameters of HTSC tokamak HTSC tokamak Major radius (m) Minor radius (m) Elongation Plasma current (MA) Toroidal field at plasma center (T) Fusion power (MW) Net energy gain
Fig. 3. Structural cross section of HTSC tokamak and layout of HTSC coils.
2.3.1. High temperature superconducting (HTSC) plasma stabilizer [4] One of the concepts is HTSC plasma stabilizer, which passively improve the plasma instability. The electromagnetic force due to type-II superconductors, to which high temperature superconductors belong, is dynamically stable because of buildup of induced current in superconductors. The stability of tokamak plasmas is attained by the induced current suppressing motion of a current carrying plasma column, as is shown in Fig. 3. The induced current is generated to keep constant the magnetic flux penetrating the HTSC coils. Advantages of the stabilizing system include the followings compared with conventional ones. It requires no power supply unlike feedback control systems. Its stabilizing effect on plasmas is free of decay, which is inevitable in that of eddy current induced in structures. Considerable plasma stabilization owing to HTSC coils provides the wider range of plasma elongation, which is usually restricted to moderate
6.0 1.88 2.1 20 6.4 650
ITER/RCO 6.2 1.9 1.67 13.3 5.51 500 10
values because of the vertical instability. In other words, HTSC coils enable compact reactors with highly elongated plasmas. Some attractive designs with use of HTSC coils were made, which displayed the feasibility of the HTSC coils for plasma stabilization and its potential to enable highly elongated plasma. This is an HTSC tokamak which is an ignition machine with the same major radius as ITER/RCO (R=6.1 m), and its cross-section and major design parameters are shown Table 2 and Fig. 3, respectively. Assumptions adopted in the design of HTSC tokamak are to use the same data base and formula as ITER in principle and not to use any innovative technology other than HTSC stabilizing coils. Employing 0-d plasma analysis, we determined major parameters such as the major radius, the aspect ratio, the elongation and the plasma current. The elongation of 2.1 is chosen to the maximum value in the range that plasma positional stability is secured with HTSC coils. Fig. 4 shows a plasma operation contour (POPCON) plot, which provides an indication of plasma operation point of Q= . To generate the contours of the auxiliary power required for plasma energy balance, the ITER-89P confinement scaling with a multiplier, H = 2.0 were employed. If we take a fusion power of 650 MW, a contour line of 650 MW gives n= 1020 m − 3 T= 14 keV as the cross point with the curve of Paux = 0. To confirm that the plasma positional stability is secured, plasma stability analysis was conducted with use of a plasma response code with
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radial and vertical directions, respectively, and the centroid moves back to the original point, followed by the recovery of bp and li. These results indicate that the plasma is stabilized by HTSC coils sufficiently, which supports for the enough plasma stabilization owing to HTSC coils and feasibility of the design of HTSC tokamak.
2.3.2. Pebble drop di6ertor [5] The pebble drop divertor is a new concept applicable to the environment of very high heat load ( \ 20 MW/m2) and particle loading (\1024 atoms per m2 s). In the pebble drop divertor system, a large number of small divertor pebbles Fig. 4. Decision of operation point in POPCON plot for HTSC tokamak.
an HTSC current analysis. Fig. 5 shows time evolution of plasma against the disturbance instantaneous bp drop of 0.20 and li drop of 0.08 recovering in 5 s. The maximum deviations of plasma centroid are −0.25 and − 0.11 m in
Fig. 5. Response of various plasma parameters during plasma disturbances.
Fig. 6. Pebble drop divertor concepts and components. (a) Multi-layer coating pebble. (b) Schematic drawing of pebble cycle.
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Fig. 7. The maximum heat flux and the surface temperature rise of divertor pebble in typical operation of the pebble drop divertor. (a) Maximum heat flux as a function of pebble radius (irradiation time, 30 ms). (b) Surface temperature rise during irradiation at incident heat flux of 30 MW/m2.
are dropped in the divertor space to form a curtain as shielding the striking zone of divertor. The pebbles are coated with at least two layers as shown in Fig. 6a, which satisfies the requirements for plasma facing material. A kernel is made from graphite or low-Z refractory ceramic like SiC, BeO or Al2O3. The mechanical strength, the thermal stress and the heat capacity of a pebble depend on the kernel. Thin plasma facing layer,
which results in low erosion character can be allowed, since the irradiation period of falling pebble is very short. Moreover, eroded pebbles can be removed at the outside the divertor. Therefore, conventional low-Z materials and coating techniques can be applied for plasma facing layer. The pebble drop divertor system consists of (1) a drop controller array, (2) pebble processing systems including a regeneration stage, a heat exchanger and the separation stage of eroded pebbles, (3) a conveying system, which is schematized in Fig. 6b. Outstanding features of the pebble drop divertor system with multi-layer pebbles and the external pebble processing systems includes, 1. pebble conveying without the MHD effect; 2. simultaneous removal of incident heat flux and particle flux; 3. both pumping function and low bulk tritium are realized by tritium permeation barrier; 4. the feasibility of divertor plasma control by mixing various pebbles with different PFL; and 5. continuous replacement of eroded divertor surface. The feasibility of the pebble drop divertor system should be verified in view of thermal and mechanical properties. Fig. 7a shows the relation between the maximum permissible heat flux on the surface of pebble and the radius of the kernel. The permissible heat flux is estimated such that the surface stress does not exceed the compressive strength of kernel material. Fig. 7b is the rise of surface temperature at the divertor heat flux of 30 MW/m2. The results imply that pebble drop divertor system with pebble of 0.5–1 mm in radius can handle the high heat flux.
3. Discussion on strategy of fusion technology development
3.1. Requirements of fusion energy commercialization Basically speaking, the effort for fusion engineering will be completed when three kinds of feasibilities, i.e. technology, safety and economics,
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are fully satisfied. In this regard, it would be reasonable to consider that technical feasibility and safety capability necessary for a future fusion reactor will be demonstrated by a success of the ITER project. This verification of the two feasibilities with a reactor like plant is really important in the long-term fusion R&D effort. Along this line, the R&D efforts have been made for the ITER technologies as the seven elements of technology development [6,7] and are close to completion. In this context, however, we are confronted with a very crucial issue regarding a viability of a fusion reactor, i.e. how we can solve the remaining issue of cost. It would therefore be right to recognize that in-depth effort of technology is required to conquer the very difficult technology for an economical commercial reactor. Purpose of the special effort for a Demo is dual. One is a possibility to change definition of the role of a Demo as demonstration of an economical reactor from the previous definition of a Demo reactor for a proof of safety capability. A remarkable benefit from the new definition is that commercialization of fusion energy gets faster by more than one decade due to omission of the effort for safety issue of a Demo. Schedules of ITER and a Demo are shown tentatively in Fig. 8. An important aspect in Fig. 8 is that a considerable technical R&D effort for a Demo will start together with ITER construction and the Demo will generate electricity continuously with high availability in 30 years after the construction. Another purpose is to re-organize structure of international collaboration, which could significantly save man power and budget leading to avoidance of double or triple investments of them to development of innovative technology for an
Fig. 8. R&D schedule for ITER and Demo reactor.
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economical fusion reactor. International task sharing should be favored as an optimum of efforts. If the ITER project will be initiated, the number of fusion researchers is supposed to decrease. To avoid this unfavorable situation, parallel paths to the commercialization of fusion energy, as is shown in Fig. 8, should be taken to realize the above two benefits so that the reduction of scale of fusion research can be avoided.
3.2. Strategy of fusion technology Next will arise an important and difficult question of how we can solve the technology issue enabling an economical reactor. In recognition of the technical difficulty of a fusion power reactor, the committee on fusion research of Science Council of Japan founded a subcommittee to deal with this issue to develop cost-effective component technology. The subcommittee discussed technical issues for cost effectiveness for 10 months since October 1998 and is finalizing a report. Essential part of the report seems important from a viewpoint of strategy for fusion technology development and is introduced in the following. In general, there is a firm belief in inherent capability of technology progress. According to the very old Japanese newspaper, it predicted 23 technical items for future technology as promising to come true 100 years later. Some of the predicted technologies realized within 100 years were a long-distance call, a jumbo jet, a shinkansen (Japanese fast train), TV, air-condition, etc. Twenty items of them were realized now to our surprise. Thus, we can believe that fusion technology necessary for a commercial reactor will be achieved in due time in view of supports from very vapid progress of technology in other fields. We can point out an example. As introduced previously, a progress of high temperature super conductor technology is very rapid and at present it is possible to fabricate a superconducting small magnet of several Tesla with HTSC tapes. If the technology of HTSC will be technically available for toroidal and poloidal coil systems in the near future, performance of a fusion reactor will be improved very much leading to a cost effective
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maintainability and operationability. Another example is of course development of excellent structural material. There is a structured order to construct a framework of the technology development for an enhanced fusion reactor. We had better speculate a strategic frame, consisting of systematization of fusion reactor technology and of a construction of a virtual fusion power plant. The virtual plant concept consists of a material virtual laboratory as a first step, next an information virtual laboratory as a second one and finally a creation virtual laboratory as an integration of them. They are shown in Fig. 9 and it is important to notice that the first level is enveloped in the second one and supports it. The first and second levels are included in the third level and utilized fully in a system integrating them.
Fig. 9. Strategic frame for systematization of fusion reactor technology.
reactor by achievement of higher magnetic field. In fact fusion power is proportional to toroidal magnetic field. Furthermore change of cryogenic coolant from liquid helium to liquid nitrogen makes a fusion system simpler leading to facile
3.2.1. Systematization of fusion technology There are several diversed notions about the definition of systematization of fusion technology so that the author would define it as follows. 1. As fundamental feature of the systematization, it takes the same pattern as the langue-parole form common in the linguistics and enables technological creation based on the systematized paradigm of fusion technology. Construction of the paradigm is a goal of our effort. 2. It is possible to predict physical fusion phenomena in materials and technical performance of a fusion system or a fusion subsystem if one consults with the systematized fusion technology paradigm. At this moment, technical R&D for ITER has been conducted for many elements of it. These were conducted to achieve targets set for the ITER technology within well-defined constraints or with boundary conditions. Unless systematization is tried, we will have to do similar things again in future when we are required to provide similar results for different technical conditions. This is a waste of time and resources and should be avoided. Consequently, technical efforts have to be made at the same time to construct theory of fusion engineering referring to experimental contents and results of the above technical R&D.
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3.2.2. De6elopment of integrated fusion technology There are three characteristics in the integrated fusion technology, i.e. (a) capability of progress, (b) structuralization and (c) integration. In Fig. 9 is shown structure of the integrated fusion technology. It consists of three levels. In the first level, various interactions between materials of components and general loadings (thermal, mechanical, particle and electromagnetic environmental ones) are of design concern and prediction capability of their behaviors should be secured to some extent of accuracy. The material –loading interactions are observed in most structural and functional components of a reactor. By constructing methods of the predictions we could consider that identification of fusion technology is established. Most of the researches conducted up to now are in this level. Still now, very important technical issues are not solved. Neutron irradiation problem of structural materials is a typical one of them. An important element in this level is to find scientific or technical rules and theory to establish ways of prediction of phenomena observed in a fusion reactor. In the second level, an information laboratory is constructed virtually, where resources of computer systems are utilized fully leading to saving up of R&D budget and speeding up of technical R&D significantly. In other words, simulation capability are applied to analysis of non-linear and complex phenomena observed in components of a real fusion machine and minimized experiments are conducted to verify validity of the simulation. If we can get sufficient agreement between two we judge that prediction capability of long-term phenomena like degradation of materials is secured and then we proceed toward a very important phase of ‘controlling phenomena’. Examples of controlling phenomena are to reduce tritium inventory and to improve strength of structural materials by some technique, to decrease heat load to a diverter plate by adding useful elements, etc. In the third level, integrated functions of a reactor are of concern. They are an assurance of long-term safety capability, remote maintenance capability, avoidance of troubles and accidents
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and finally achievement of economical requirements. When we consider how these integrated functions of a reactor are satisfied, we are led to an introduction of a virtual plant concept in the strategic frame of the systematized research plan. The virtual plant proposed here is different from a conventional concept in two points. One is an introduction of a real time in the virtual plant. This is possible by introduction of ‘degradation of materials’ used in a plant. And another is the possibility to visualize major fusion phenomena inside a reactor and situations of reactor components at any time using systematic simulation tools and other means. Coincidence between a real plant and a virtual one should be secured at necessary time intervals using measurement data set of the real plant, which is a necessary input to various simulation codes. Once such a virtual plant is constructed, we can conduct virtual experiments arbitrarily with the model by changing design parameters and observing reactor responses in expectation of new technical findings and other similar things. In order to introduce a real time in a reactor, we need two different informations on degradation behaviors of components with time and history of loading conditions and environment where components of concern are subjected. If the two informations are available fully, it might become possible to predict any accidents to some extent of accuracy. This is of course a very tough task but we are required to challenge this goal. Economical improvement is expected possibly by a simulation study with the virtual plant through forward and feed back manipulation where significance of a new technology is also evaluated as to contribution, cost reduction and high performance of a reactor. At the same time R&D requirements are manifested and will be carried out at a proper organization and by organized researchers.
4. Conclusion 1. Up to now administration and organization of fusion research in Japan have been conducted under auspices of Monbusho and STA.
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2. However, both will be unified in 2001 and restructuring of the fusion research system will be discussed and different shapes of it might appear, according to a progress of the situation. 3. Steady progress of fusion technology at university has been made aiming at the systematization of it for long-term effort. 4. For restructuring of fusion technology researches, the new concept of integrated fusion technology is defined and proposed based on the discussions at the committee on Fusion Research, Science Council of Japan through associated systematization of fusion technology. 5. Development of a fusion virtual plant is an excellent approach and must be different from a conventional concept of a virtual plant in introduction of real time and visualization of working state of the plant at any time. Real time can be realized through an introduction of degradation parameter of component and material. Acknowledgements The author would like to express his sincere appreciation to a contribution by Dr T. Uchimoto
.
to the paper. Also a useful discussion with Professor A. Kohyama and Professor N. Sekimura about fusion material should be noted here. Professor Imagawa and Professor Motojima should also be noted here for providing materials on the helical coils.
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