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Status of fusion technology development in JAERI stressing steady-state operation for future reactors
Fusion Engineering and Design 49 – 50 (2000) 27 – 32 www.elsevier.com/locate/fusengdes
Status of fusion technology development in JAERI stressing steady-state operation for future reactors Shinzaburo Matsuda * Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801 -1 Mukoyama, Naka-machi, Naka-gun, Ibaraki-ken, 311 -0193, Japan
Abstract This paper reports on the progress of the fusion reactor technologies developed at the Japan Atomic Energy Research Institute (JAERI) and expected to lead to a future steady state operation reactor. In particular, superconducting coil technology for plasma confinement, NBI and RF systems technology for plasma control and current drive, fueling and pumping systems technology for particle control, heat removal technology, and development of long life materials are highlighted as the important key elements for the future steady state operation. It will be discussed how these key technologies have already been developed by the ITER (International Thermonuclear Experimental Reactor) technology R&D as well as by the Japanese domestic program, and which technologies are planned for the near future. © 2000 Elsevier Science B.V. All rights reserved. Keywords: JAERI; Fusion reactor technologies; Superconducting coil technology
1. Introduction Various reactor designs and their assessments as well as the promising data obtained in recent years in tokamaks have brought fusion communities and governments to the widely-shared opinion that the future goal of energy production by toroidal magnetic fusion will be reached through the high beta (energy density) steady state reactor. Such a perspective has led to emphasize studies of a steady state DEMO reactor and to modify the detailed ITER technical objectives and designs. * Tel.: +81-29-2707500; fax: + 81-29-2707519. E-mail address: [email protected] (S. Matsuda).
Thus, it is important to address and develop fusion reactor technologies by stressing this goal. In 1992 the Atomic Energy Commission (AEC) of Japan appointed JAERI as the central institute responsible for the development of an experimental reactor, a core machine in the third phase national fusion R&D program. It’s task covers the research, development and construction of the experimental reactor. Since the scope and the technical objectives of ITER are almost identical with those of the Japanese third phase program, the AEC has encouraged JAERI to become more and more actively involved in the ITER activities. Within the ITER framework, ITER technology R&D has been internationally conducted on a large scale. Particularly in Japan, a variety of
0920-3796/00/$ - see front matter © 2000 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 - 3 7 9 6 ( 0 0 ) 0 0 4 0 1 - 4
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S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
R&D activities have been conducted in close collaboration with national institutes, many universities and more than 30 top level industries. These R&D activities are the basic technologies essential for the future reactors. However, from the socioeconomic point of view on attractive future fusion reactors, cost and environmental impact should be emphasized in reactor designs and associated research and developments. The implications of the study on the economical reactor design carried out by Kikuchi et al. [1,2] and others [3] have shown a primary need for higher beta plasma confinement, higher magnetic field generation and a longer time between regular tritium breeding blanket replacement. The study by Seki et al. [4] has indicated that use of low activation materials such as ferritic steel F82H as a reactor structural material will meet the basic requirements. Vanadium Alloy or SiC/SiC composite are considered to increase attractiveness, should the design prove to be valid by means of the R&D programme. JAERI has been making a strong effort in both design and technology R&D for the steady state DEMO reactor and future commercial reactors. 2. Organization for fusion technology at JAERI At JAERI, the Naka Fusion Research Establishment is responsible for covering most of the
Fig. 1. The magnetic field strength at the conductor position versus the conductor current. Comparison was made of the past achievements by the world SC magnets with the ITER central solenoid and toroidal field magnets.
fusion activities. Since last April, two laboratories on materials development and fusion neutronics studies have been moved into Naka administration from Tokai. JAERI has wide-range close collaborations with universities and national institutes in the framework of a JAERI universities/ national institute collaboration scheme. JAERI plays a central role in the implementation of R&D. Many industries are involved in R&D activities through procurement and fabrication contracts and through secondment of engineers to JAERI. With the progress of fusion technology R&D, deeper and wider involvement of industries will be needed because of the increasing role of manufacturing and fabrication technology. Therefore, some R&D activities have to be transferred to these industries.
3. Superconducting magnet The superconducting magnets are one of the most important elements for a fusion reactor, since they play a significant role in terms of performance and cost. For instance, one third of the direct construction cost of ITER is due to the manufacturing of superconducting magnets. Through the ITER engineering activities, significant progress has been made in the development and the fabrication technology of large volume, strong field magnets for both DC toroidal and pulse poloidal magnetic field generation. Fig. 1 shows the comparison of the past achievements by the world superconducting magnets with the present ITER targets in terms of conductor current and magnetic field strength at the conductor. Further progress will have to be made with the development of winding technology for even large volume, stronger magnetic field generation and concomitant higher stress coils. Requirements for DEMO and later reactor magnets will be about 16.5 T, 80 kA. It is clear however, that ITER can provide stepped-up technologies for the future fusion reactors in both DC and pulsed magnets. The ITER Central Solenoid Model Coil aims at verifying, on a reduced scale model, the fundamental performance and manufacturing feasibility
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
29
Fig. 2. Outer module of the CS model coil for ITER before installation into the test chamber. The coil is 3.6 m diam., 2.8 m high, and about 100 tons in total.
of the ITER CS coil. It can be operated at a conductor current of 46 kA by generating a magnetic field of 13 T, and the rate of field change of 1 T/s [5]. It is composed of 18 layered windings and a single layer insert both using Nb3Sn as a superconducting material (Fig. 2). This Model Coil was jointly developed and fabricated in collaboration with US, EU, RF home teams. JAERI is the host institute for this project and had
successfully contributed to the reduction of the AC heat loss by a factor of five in Nb3Sn strand in order to match the ITER requirement. The subsequent production of 13.6 tons of strands, and their cabling with high QA and QC has also contributed to establish a firm fabrication technology. In the EU industry, the strands were then inserted into the armour conduit specifically developed for Nb3Sn by the US, and compacted to
30
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
form a conductor. The conductor drums thus manufactured were sent back to US and Japan for coil windings of the Inner and the Outer Module, respectively. After successful fabrication of these Modules and Insert, they were shipped to the JAERI Naka site where the installation into the vacuum chamber, 6.5 m in diameter and 9.5 m high, has been underway since June 1999. The test facility for the CS model coil was completed in 1996, and has been waiting for coil installation. The main feature of this system is the capability of 50 kA current at 4.5 kV using JT-60 pulsed power supply. The rated refrigerator capacity is 5 kW at 4.5 K. One of the figures of merit representing the refrigeration efficiency is the ratio of the refrigeration capacity divided by the electric power consumed. The refrigerator of this test facility consumes about 1.8 MW electric power equivalent, giving an efficiency number of 1/360 instead of foregoing efficiency number of 1/500 –1/700. This can be further improved by employing an additional expansion turbine for supercritical helium in the refrigeration process, thus leading to a designed refrigeration efficiency of 1/250–1/300 for the ITER refrigerator of 70 kW at 4.5 K. To achieve a stronger magnetic field in DEMO reactors, one of the attractive materials for superconducting coils is Nb3Al. As an effort for an alternative candidate for ITER TF coil, and as a national effort, development of a single layered Nb3Al insert coil, 1.4 m diam. and 2.8 m high is underway. After successful strand development, production of strands, cabling and packing into the circular stainless steel thin conduit to form a conductor has been done. The winding procedure to form a coil is now underway. In parallel to this first step comprehensive approach, a target was set to increase the critical current density by a factor of three. To this purpose further development of fabrication technologies is needed to utilize the inherent characteristics of the Nb3Al material more. 4. Plasma control and current drive Heating of the plasma into a sufficiently high Q-value regime is primarily needed for the toka-
mak systems, and can be achieved by various methods, but this does not require continuous operation. However, non-inductive current drive for steady state operation may not allow many options. The neutral beam injection, using negative deuterium ion as primary beam, has become one of the most promising methods. The main target performances required for the NBI system for the DEMO reactor will be a steady state operation of the negative ion source, high energy acceleration, maintenance free ion source, a high system power efficiency, a steady state cryopump and the radiation hard performance. After achieving 1 MeV acceleration, negative H-ions of a high current density of 30 mA/cm2 have been extracted from the ‘‘KAMABOKO’’ source developed by JAERI producing a total of 140 mA [6]. For the steady state operation, the source can be operated more than 140 h at a current density exceeding 10 mA/cm2. As a maintenance free ion source development, an ion source without filaments using 2.45 GHz microwave and water cooled ceramic windows is under testing and has demonstrated the possibility of steady state operation. The dominant element for the overall power efficiency is the neutralization efficiency of about 60% for the gas neutralizer, but this can be further increased to about 85% by application of the plasma neutralizer. This effort has been jointly undertaken with Kurchatov Institute, Kyoto University and will be tested at the JAERI test stand in the year 2000. In addition, tests of a continuously operational cryopump are underway with a pumping speed of 50 m3/s. The ITER prototype gyrotron, which has TE31,8 cavity and CVD diamond window has achieved 520 kW/6 s and 450 kW/8 s operation at 170 GHz. The CVD diamond has outstanding performance as a mm wave window. The RF loss is 1/10 of the sapphire and the thermal conductivity is 40 times higher than that of sapphire. The first results of the CVD diamond window gyrotron have shown the diamond window’s capability to deliver more than 1 MW at 170 GHz in CW [7]. Fig. 3 shows measured and calculated temperature evolutions of the window for two cases of
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
transmission performance. Thus, the window issue has been solved for the ITER gyrotron. A stable oscillation of more than 1 MW at the cavity had already been obtained. The most critical issue of the ITER gyrotron left for CW operation is the unexpected reduction in the RF mode generation. There are several reasons. One is the parasitic oscillation around the cavity. Another one is the mode conversion to the lower order modes at the downstream side after cavity. Diffraction at the mode converter and reflection from several junctions are also causes of the unexpected modes. These phenomena are now being studied and they will be clarified and solved during the EDA extension period. The frequency of the ECRF system in DEMO will be substantially higher than that of ITER. More than 300 GHz is required in the DEMO corresponding to the higher central magnetic field. The window will not be critical because the diamond window still has more than 1 MW capability at this frequency. The most critical issue is the cavity. The new cavity, which has good stability in super high order mode, must be developed. The co-axial cavity and the self-mode selective cylindrical cavity are the candidates for this high frequency operation in CW.
5. Heat removal technology Tritium breeding blankets have to be developed for continuous energy extraction and continuous
Fig. 3. Time dependence of window temperature. Closed circles are experimental data at 0.5 MW power transmission. Dashed and solid lines are simulation results for two cases of transmission performance.
31
tritium production. These blankets will be developed by a combination of the tests, one through the blanket module tests using the ITER as volumetric neutron environment, and the other one through irradiation tests of the small low activation materials at the 14 MeV neutron source and the fission reactors. On the grounds of both socioeconomic and thermal characteristics for high temperature, we consider the ferritic steel F82H to be the most suitable for the basic material for the DEMO and the commercial plants. In this line of development, a first wall model for DEMO breeding blanket made of ferritic steel F82H was fabricated. It is a 120 mm wide, 200 mm long and 18 mm thick flat plate which has 10 rectangular cooling channels fabricated by the Hot Isostatic Pressing bonding method. The model was tested and endured 5000 cyclic surface heat load of 3 MW/m2, at the temperature of about 450°C. Details of this component development will be presented by Dr M. Seki at this Symposium [8]. The divertor requires most frequent maintenance, since the divertor plates are exposed to the highest heat and particle fluxes and also exposed to neutron irradiation. Choice of the armour block and the cooling tube materials for the divertor structure is critical for the steady state reactor. This will be dependent on the level of the heat load to the divertor plate. The heat load condition, although very important to the design and R&D for DEMO reactors, is of a wide range now, and will be narrowed down, but remain unfixed until the final confirmation by ITER, depending on the effectiveness of the radiation cooling by the divertor plasma. In case the heat load to the divertor is less than about 5 MW/m2, Tungsten armoured ferritic steel cooling tube will be a suitable choice for the design, and the regular interval maintenance will be necessary. The frequency of the replacement is determined by the thermal fatigue life and creep life of ferritic steel F82H, and the irradiation embrittlement of the tungsten armour against neutrons. In the case of higher heat loading, dispersion strengthened tungsten armour and ferritic steel have to be developed. So far, the R&D for DEMO reactor has been performed at a very limited level. Basic performance tests on the ferritic steel F82H, such
32
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
as thermal fatigue life test and the critical heat flux test, have started. Details of this development will be presented at this symposium [8].
reasonable confidence that we will obtain sufficient technological information to step into realization of the DEMO reactor.
6. Fueling and pumping technology
References
As to the steady state pumping systems, an oil-free mechanical pump with helical grooves on the surface of the rooter operable up to 1000 Pa hydrogen outlet pressure has been developed. A combination of this pump with commercially available turbo-molecular pump will provide continuous pumping for the pressure range of 10 − 7 – 10 − 1 Pa range. Further, through the introduction of the gas separation system under development, helium gas can be separated selectively at room temperature, and the residual DT gas will be directly returned to the fueling system.
[1] Y. Seki, M. Kikuchi, The steady state tokamak reactor, 13th IAEA Conf., Washington, 1990, IAEA-CN-53/G-1-2. [2] M. Kikuchi, Y. Seki, K. Nakagawa, The Advanced SSTR, 6th IAEA TCM on Fusion Power Plant Design and Technology, Culham, March 1998 (to be published in Fusion Engineering and Design). [3] ARIES Team, The STARLITE study, assessment of options for tokamak power plants, final report, UCSD-ENG005, 1997. [4] Y. Seki, T. Tabara, I. Aoki, S. Ueda, S. Nishio, R. Kurihara, Impact of low activation materials on fusion reactor design, J. Nucl. Mater. 258 – 263 (1998) 1791 –1797. [5] H. Tsuji, T. Ando, D. Bessette, E. Egorov, T. Isono, R. Jayakumar, T. Kato, N. Martovetsky, K. Okuno, S. Shimamoto, R. Thome, R. Vieira, J. Wohlwend, ITER Central Solenoid Model Coil Test Program, 17th IAEA Conf., Yokohama, 1998, IAEA-CN-69/ITERP1/22. [6] M. Hanada, N. Akino, N. Ebisawa, Y. Fujiwara, A. Honda, T. Itoh, M. Kawai, M. Kazawa, M. Kuriyama, K. Mogaki, Y. Okumura, H. Oohara, K. Oomori, K. Usui, K. Watanabe, Development of Multi-Mega watt negative ion sources and accelerators for neutral beam injectors, 17th IAEA Fusion Energy Conf., Yokohama, 1998, IAEA-CN69-FTP/20. [7] K. Sakamoto, A. Kasuga, M. Tsuneoka, K. Takahashi, T. Imai, T. Kariya, Y. Mitsunaka, Rev. Sci. Instrum. 70 (1999) 208. [8] M. Seki, Development and testing of large scale nuclear components and remote handling equipment in JAERI, Fusion Engineering and Design (invited symposium paper).
7. Summary Status of the reactor technology R&D in JAERI for the steady state DEMO and later reactors has been reviewed. Emphasis has been placed on the development of the elements for achieving steady state, since significant parts of the common R&D can be achieved by ITER. Although the level of the budgetary investment for these activities is so far modest, the achievements are steady and promising. Therefore, assuming that ITER will be built, there is a
.
Status of fusion technology development in JAERI stressing steady-state operation for future reactors Shinzaburo Matsuda * Department of Fusion Engineering Research, Naka Fusion Research Establishment, Japan Atomic Energy Research Institute, 801 -1 Mukoyama, Naka-machi, Naka-gun, Ibaraki-ken, 311 -0193, Japan
Abstract This paper reports on the progress of the fusion reactor technologies developed at the Japan Atomic Energy Research Institute (JAERI) and expected to lead to a future steady state operation reactor. In particular, superconducting coil technology for plasma confinement, NBI and RF systems technology for plasma control and current drive, fueling and pumping systems technology for particle control, heat removal technology, and development of long life materials are highlighted as the important key elements for the future steady state operation. It will be discussed how these key technologies have already been developed by the ITER (International Thermonuclear Experimental Reactor) technology R&D as well as by the Japanese domestic program, and which technologies are planned for the near future. © 2000 Elsevier Science B.V. All rights reserved. Keywords: JAERI; Fusion reactor technologies; Superconducting coil technology
1. Introduction Various reactor designs and their assessments as well as the promising data obtained in recent years in tokamaks have brought fusion communities and governments to the widely-shared opinion that the future goal of energy production by toroidal magnetic fusion will be reached through the high beta (energy density) steady state reactor. Such a perspective has led to emphasize studies of a steady state DEMO reactor and to modify the detailed ITER technical objectives and designs. * Tel.: +81-29-2707500; fax: + 81-29-2707519. E-mail address: [email protected] (S. Matsuda).
Thus, it is important to address and develop fusion reactor technologies by stressing this goal. In 1992 the Atomic Energy Commission (AEC) of Japan appointed JAERI as the central institute responsible for the development of an experimental reactor, a core machine in the third phase national fusion R&D program. It’s task covers the research, development and construction of the experimental reactor. Since the scope and the technical objectives of ITER are almost identical with those of the Japanese third phase program, the AEC has encouraged JAERI to become more and more actively involved in the ITER activities. Within the ITER framework, ITER technology R&D has been internationally conducted on a large scale. Particularly in Japan, a variety of
0920-3796/00/$ - see front matter © 2000 Elsevier Science B.V. All rights reserved. PII: S 0 9 2 0 - 3 7 9 6 ( 0 0 ) 0 0 4 0 1 - 4
28
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
R&D activities have been conducted in close collaboration with national institutes, many universities and more than 30 top level industries. These R&D activities are the basic technologies essential for the future reactors. However, from the socioeconomic point of view on attractive future fusion reactors, cost and environmental impact should be emphasized in reactor designs and associated research and developments. The implications of the study on the economical reactor design carried out by Kikuchi et al. [1,2] and others [3] have shown a primary need for higher beta plasma confinement, higher magnetic field generation and a longer time between regular tritium breeding blanket replacement. The study by Seki et al. [4] has indicated that use of low activation materials such as ferritic steel F82H as a reactor structural material will meet the basic requirements. Vanadium Alloy or SiC/SiC composite are considered to increase attractiveness, should the design prove to be valid by means of the R&D programme. JAERI has been making a strong effort in both design and technology R&D for the steady state DEMO reactor and future commercial reactors. 2. Organization for fusion technology at JAERI At JAERI, the Naka Fusion Research Establishment is responsible for covering most of the
Fig. 1. The magnetic field strength at the conductor position versus the conductor current. Comparison was made of the past achievements by the world SC magnets with the ITER central solenoid and toroidal field magnets.
fusion activities. Since last April, two laboratories on materials development and fusion neutronics studies have been moved into Naka administration from Tokai. JAERI has wide-range close collaborations with universities and national institutes in the framework of a JAERI universities/ national institute collaboration scheme. JAERI plays a central role in the implementation of R&D. Many industries are involved in R&D activities through procurement and fabrication contracts and through secondment of engineers to JAERI. With the progress of fusion technology R&D, deeper and wider involvement of industries will be needed because of the increasing role of manufacturing and fabrication technology. Therefore, some R&D activities have to be transferred to these industries.
3. Superconducting magnet The superconducting magnets are one of the most important elements for a fusion reactor, since they play a significant role in terms of performance and cost. For instance, one third of the direct construction cost of ITER is due to the manufacturing of superconducting magnets. Through the ITER engineering activities, significant progress has been made in the development and the fabrication technology of large volume, strong field magnets for both DC toroidal and pulse poloidal magnetic field generation. Fig. 1 shows the comparison of the past achievements by the world superconducting magnets with the present ITER targets in terms of conductor current and magnetic field strength at the conductor. Further progress will have to be made with the development of winding technology for even large volume, stronger magnetic field generation and concomitant higher stress coils. Requirements for DEMO and later reactor magnets will be about 16.5 T, 80 kA. It is clear however, that ITER can provide stepped-up technologies for the future fusion reactors in both DC and pulsed magnets. The ITER Central Solenoid Model Coil aims at verifying, on a reduced scale model, the fundamental performance and manufacturing feasibility
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
29
Fig. 2. Outer module of the CS model coil for ITER before installation into the test chamber. The coil is 3.6 m diam., 2.8 m high, and about 100 tons in total.
of the ITER CS coil. It can be operated at a conductor current of 46 kA by generating a magnetic field of 13 T, and the rate of field change of 1 T/s [5]. It is composed of 18 layered windings and a single layer insert both using Nb3Sn as a superconducting material (Fig. 2). This Model Coil was jointly developed and fabricated in collaboration with US, EU, RF home teams. JAERI is the host institute for this project and had
successfully contributed to the reduction of the AC heat loss by a factor of five in Nb3Sn strand in order to match the ITER requirement. The subsequent production of 13.6 tons of strands, and their cabling with high QA and QC has also contributed to establish a firm fabrication technology. In the EU industry, the strands were then inserted into the armour conduit specifically developed for Nb3Sn by the US, and compacted to
30
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
form a conductor. The conductor drums thus manufactured were sent back to US and Japan for coil windings of the Inner and the Outer Module, respectively. After successful fabrication of these Modules and Insert, they were shipped to the JAERI Naka site where the installation into the vacuum chamber, 6.5 m in diameter and 9.5 m high, has been underway since June 1999. The test facility for the CS model coil was completed in 1996, and has been waiting for coil installation. The main feature of this system is the capability of 50 kA current at 4.5 kV using JT-60 pulsed power supply. The rated refrigerator capacity is 5 kW at 4.5 K. One of the figures of merit representing the refrigeration efficiency is the ratio of the refrigeration capacity divided by the electric power consumed. The refrigerator of this test facility consumes about 1.8 MW electric power equivalent, giving an efficiency number of 1/360 instead of foregoing efficiency number of 1/500 –1/700. This can be further improved by employing an additional expansion turbine for supercritical helium in the refrigeration process, thus leading to a designed refrigeration efficiency of 1/250–1/300 for the ITER refrigerator of 70 kW at 4.5 K. To achieve a stronger magnetic field in DEMO reactors, one of the attractive materials for superconducting coils is Nb3Al. As an effort for an alternative candidate for ITER TF coil, and as a national effort, development of a single layered Nb3Al insert coil, 1.4 m diam. and 2.8 m high is underway. After successful strand development, production of strands, cabling and packing into the circular stainless steel thin conduit to form a conductor has been done. The winding procedure to form a coil is now underway. In parallel to this first step comprehensive approach, a target was set to increase the critical current density by a factor of three. To this purpose further development of fabrication technologies is needed to utilize the inherent characteristics of the Nb3Al material more. 4. Plasma control and current drive Heating of the plasma into a sufficiently high Q-value regime is primarily needed for the toka-
mak systems, and can be achieved by various methods, but this does not require continuous operation. However, non-inductive current drive for steady state operation may not allow many options. The neutral beam injection, using negative deuterium ion as primary beam, has become one of the most promising methods. The main target performances required for the NBI system for the DEMO reactor will be a steady state operation of the negative ion source, high energy acceleration, maintenance free ion source, a high system power efficiency, a steady state cryopump and the radiation hard performance. After achieving 1 MeV acceleration, negative H-ions of a high current density of 30 mA/cm2 have been extracted from the ‘‘KAMABOKO’’ source developed by JAERI producing a total of 140 mA [6]. For the steady state operation, the source can be operated more than 140 h at a current density exceeding 10 mA/cm2. As a maintenance free ion source development, an ion source without filaments using 2.45 GHz microwave and water cooled ceramic windows is under testing and has demonstrated the possibility of steady state operation. The dominant element for the overall power efficiency is the neutralization efficiency of about 60% for the gas neutralizer, but this can be further increased to about 85% by application of the plasma neutralizer. This effort has been jointly undertaken with Kurchatov Institute, Kyoto University and will be tested at the JAERI test stand in the year 2000. In addition, tests of a continuously operational cryopump are underway with a pumping speed of 50 m3/s. The ITER prototype gyrotron, which has TE31,8 cavity and CVD diamond window has achieved 520 kW/6 s and 450 kW/8 s operation at 170 GHz. The CVD diamond has outstanding performance as a mm wave window. The RF loss is 1/10 of the sapphire and the thermal conductivity is 40 times higher than that of sapphire. The first results of the CVD diamond window gyrotron have shown the diamond window’s capability to deliver more than 1 MW at 170 GHz in CW [7]. Fig. 3 shows measured and calculated temperature evolutions of the window for two cases of
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
transmission performance. Thus, the window issue has been solved for the ITER gyrotron. A stable oscillation of more than 1 MW at the cavity had already been obtained. The most critical issue of the ITER gyrotron left for CW operation is the unexpected reduction in the RF mode generation. There are several reasons. One is the parasitic oscillation around the cavity. Another one is the mode conversion to the lower order modes at the downstream side after cavity. Diffraction at the mode converter and reflection from several junctions are also causes of the unexpected modes. These phenomena are now being studied and they will be clarified and solved during the EDA extension period. The frequency of the ECRF system in DEMO will be substantially higher than that of ITER. More than 300 GHz is required in the DEMO corresponding to the higher central magnetic field. The window will not be critical because the diamond window still has more than 1 MW capability at this frequency. The most critical issue is the cavity. The new cavity, which has good stability in super high order mode, must be developed. The co-axial cavity and the self-mode selective cylindrical cavity are the candidates for this high frequency operation in CW.
5. Heat removal technology Tritium breeding blankets have to be developed for continuous energy extraction and continuous
Fig. 3. Time dependence of window temperature. Closed circles are experimental data at 0.5 MW power transmission. Dashed and solid lines are simulation results for two cases of transmission performance.
31
tritium production. These blankets will be developed by a combination of the tests, one through the blanket module tests using the ITER as volumetric neutron environment, and the other one through irradiation tests of the small low activation materials at the 14 MeV neutron source and the fission reactors. On the grounds of both socioeconomic and thermal characteristics for high temperature, we consider the ferritic steel F82H to be the most suitable for the basic material for the DEMO and the commercial plants. In this line of development, a first wall model for DEMO breeding blanket made of ferritic steel F82H was fabricated. It is a 120 mm wide, 200 mm long and 18 mm thick flat plate which has 10 rectangular cooling channels fabricated by the Hot Isostatic Pressing bonding method. The model was tested and endured 5000 cyclic surface heat load of 3 MW/m2, at the temperature of about 450°C. Details of this component development will be presented by Dr M. Seki at this Symposium [8]. The divertor requires most frequent maintenance, since the divertor plates are exposed to the highest heat and particle fluxes and also exposed to neutron irradiation. Choice of the armour block and the cooling tube materials for the divertor structure is critical for the steady state reactor. This will be dependent on the level of the heat load to the divertor plate. The heat load condition, although very important to the design and R&D for DEMO reactors, is of a wide range now, and will be narrowed down, but remain unfixed until the final confirmation by ITER, depending on the effectiveness of the radiation cooling by the divertor plasma. In case the heat load to the divertor is less than about 5 MW/m2, Tungsten armoured ferritic steel cooling tube will be a suitable choice for the design, and the regular interval maintenance will be necessary. The frequency of the replacement is determined by the thermal fatigue life and creep life of ferritic steel F82H, and the irradiation embrittlement of the tungsten armour against neutrons. In the case of higher heat loading, dispersion strengthened tungsten armour and ferritic steel have to be developed. So far, the R&D for DEMO reactor has been performed at a very limited level. Basic performance tests on the ferritic steel F82H, such
32
S. Matsuda / Fusion Engineering and Design 49–50 (2000) 27–32
as thermal fatigue life test and the critical heat flux test, have started. Details of this development will be presented at this symposium [8].
reasonable confidence that we will obtain sufficient technological information to step into realization of the DEMO reactor.
6. Fueling and pumping technology
References
As to the steady state pumping systems, an oil-free mechanical pump with helical grooves on the surface of the rooter operable up to 1000 Pa hydrogen outlet pressure has been developed. A combination of this pump with commercially available turbo-molecular pump will provide continuous pumping for the pressure range of 10 − 7 – 10 − 1 Pa range. Further, through the introduction of the gas separation system under development, helium gas can be separated selectively at room temperature, and the residual DT gas will be directly returned to the fueling system.
[1] Y. Seki, M. Kikuchi, The steady state tokamak reactor, 13th IAEA Conf., Washington, 1990, IAEA-CN-53/G-1-2. [2] M. Kikuchi, Y. Seki, K. Nakagawa, The Advanced SSTR, 6th IAEA TCM on Fusion Power Plant Design and Technology, Culham, March 1998 (to be published in Fusion Engineering and Design). [3] ARIES Team, The STARLITE study, assessment of options for tokamak power plants, final report, UCSD-ENG005, 1997. [4] Y. Seki, T. Tabara, I. Aoki, S. Ueda, S. Nishio, R. Kurihara, Impact of low activation materials on fusion reactor design, J. Nucl. Mater. 258 – 263 (1998) 1791 –1797. [5] H. Tsuji, T. Ando, D. Bessette, E. Egorov, T. Isono, R. Jayakumar, T. Kato, N. Martovetsky, K. Okuno, S. Shimamoto, R. Thome, R. Vieira, J. Wohlwend, ITER Central Solenoid Model Coil Test Program, 17th IAEA Conf., Yokohama, 1998, IAEA-CN-69/ITERP1/22. [6] M. Hanada, N. Akino, N. Ebisawa, Y. Fujiwara, A. Honda, T. Itoh, M. Kawai, M. Kazawa, M. Kuriyama, K. Mogaki, Y. Okumura, H. Oohara, K. Oomori, K. Usui, K. Watanabe, Development of Multi-Mega watt negative ion sources and accelerators for neutral beam injectors, 17th IAEA Fusion Energy Conf., Yokohama, 1998, IAEA-CN69-FTP/20. [7] K. Sakamoto, A. Kasuga, M. Tsuneoka, K. Takahashi, T. Imai, T. Kariya, Y. Mitsunaka, Rev. Sci. Instrum. 70 (1999) 208. [8] M. Seki, Development and testing of large scale nuclear components and remote handling equipment in JAERI, Fusion Engineering and Design (invited symposium paper).
7. Summary Status of the reactor technology R&D in JAERI for the steady state DEMO and later reactors has been reviewed. Emphasis has been placed on the development of the elements for achieving steady state, since significant parts of the common R&D can be achieved by ITER. Although the level of the budgetary investment for these activities is so far modest, the achievements are steady and promising. Therefore, assuming that ITER will be built, there is a
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