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Watanabe, Kazuhito; Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Uto, Hiroyasu; Sakamoto, Yoshiteru; Araki, Takao*; Asano, Shiro*; Asano, Kazuhito*
Proceedings of 26th IEEE Symposium on Fusion Engineering (SOFE 2015), 6 Pages, 2016/06
Safety studies of a water-cooled fusion DEMO reactor have been performed. In the event of the blanket cooling pipe break outside the vacuum vessel, i.e. ex-vacuum vessel loss of coolant accident (ex-VV LOCA), the pressurized steam and air may lead to damage reactor building walls which have confinement function, and to release the radioactive materials to the environment. In response to this accident, we proposed three cases of confinement strategies. In each case, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to outside the boundaries were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries.
Uto, Hiroyasu; Takase, Haruhiko; Sakamoto, Yoshiteru; Tobita, Kenji; Mori, Kazuo; Kudo, Tatsuya; Someya, Yoji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; et al.
Fusion Engineering and Design, 103, p.93 - 97, 2016/02
Times Cited Count：8 Percentile：60.26(Nuclear Science & Technology)Conceptual design of in-vessel component including conducting shell has been investigated in Broader Approach (BA) DEMO design activities, in order to propose feasible DEMO reactor from plasma vertical stability and engineering viewpoint. The conducting shell for the plasma vertical stability will be incorporated behind blanket module, while the location must be close to the plasma surface as possible for the plasma stabilization. We evaluated dependence of the plasma vertical stability on the conducing shell parameters by using a 3-dimensional eddy current analysis code (EDDYCAL). The calculation results showed that the conducting shell requires more than 0.01 m thickness of Cu-alloy on DEMO. On the other hand, the electromagnetic force at the plasma disruption is a few times larger than no conducting shell case because of larger eddy current on conducting shell. The engineering design issues of in-vessel components for plasma vertical stability are presented.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Uto, Hiroyasu; Sakamoto, Yoshiteru; Gulden, W.*
Nuclear Fusion, 55(12), p.123008_1 - 123008_7, 2015/12
Times Cited Count：15 Percentile：60.5(Physics, Fluids & Plasmas)Major in- and ex-vessel loss-of-coolant accidents (LOCAs) of a water-cooled tokamak fusion DEMO reactor have been analysed. Analyses have identified responses of the DEMO systems to these accidents and pressure loads to confinement barriers for radioactive materials. The thermohydraulic analysis results suggests that the in- and ex-vessel LOCAs crucially threaten integrity of the primary and final confinement barriers, respectively. As for the in-vessel LOCA, it was found that the pressure in the vacuum vessel reaches its design value due to the LOCA even though a pressure suppression system is in service. As for the ex-vessel LOCA, the pressure load to the tokamak hall due to the double-ended break of the primary cooling pipe was found to be so large that integrity of the hall was crucially challenged. Mitigations of the loads to the confinement barriers are also discussed.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Tokunaga, Shinsuke; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Sakamoto, Yoshiteru
Fusion Engineering and Design, 98-99, p.1872 - 1875, 2015/10
Times Cited Count：42 Percentile：96.3(Nuclear Science & Technology)Blanket concept with simplified interior for mass production has been developed with a mixed bed of Li
TiO
and Be
Ti pebbles, a coolant condition of 15.5 MPa and 290-325
C and cooling tubes only without any partitions. A neutronics analysis ensured the blanket concept meets a self-sufficient supply of tritium. However, this concept is vulnerable to the inner pressure. A plant availability for DEMO may drop to a lower value, because a potential of resume operations after an accident such as a coolant leakage in blanket is not considered. The blanket design will be revisited for the availability. Considering the continuity with the ITER-TBM option of Japan and the engineering feasibility of fabrication, our design study focuses on a water-cooled solid breeding blanket using the mixed pebbles bed. A breakage of the blanket casing should be avoided not to contaminate the plasma chamber with water and breeding materials. A water-cooled solid blanket with inner pressure tightness is estimated by the ANSYS code. As a results, the pressure tightness of 8 MPa (water vapor pressure at 300C) can be compatible with the self-sufficient production of tritium when the blanket is as thick as about 0.9 m and the ribs are arranged in the radial direction. Therefore, the blanket concept with pressure tightness of 8 MPa is adopted with depressurization system as which a tritium recovery system such as helium purge-gas line is posteriorly arranged in blanket to serve. On the other hand, a handling of decay heat is a serious problem at an accident such as LOCA. Coolant flow is divided into the blanket to secure heat removal for the safety. Finally, the blanket segmentation with the shape and dimension of blanket and routing of coolant flow has also been proposed. Moreover, overall TBR is estimated with torus configuration based in the segmentation using three-dimensional MCNP calculation.
Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto
Fusion Engineering and Design, 98-99, p.1648 - 1651, 2015/10
Times Cited Count：7 Percentile：51.25(Nuclear Science & Technology)Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field coil, the arrangement of poloidal field coils, reactor building, hot cell and so forth. Therefore, the maintenance schemes should satisfy many design requirements and criteria to assure reliable and safe plant operation and to attain reasonable plant availability. In this study, we categorize various schemes in term of (1) the maintenance port position for transporting blanket segments, (2) blanket segmentation, and (3) divertor segmentation. In reviewing these assessment factors, the separated sector transport using the vertical maintenance ports with small divertor cassette maintenance scheme was found to be a more probable maintenance approach. This presentation describes engineering design of each maintenance schemes and evaluation results of comparison among maintenance schemes.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
Fusion Science and Technology, 68(2), p.423 - 427, 2015/09
Times Cited Count：12 Percentile：70.51(Nuclear Science & Technology)The radioactive waste is generated in every replacement of an in-vessel component. Maintenance scheme is to replace the blanket segment and divertor cassette independently, as the lifetime of them is different. The blanket segment consists of some blanket modules mounted to back-plate. Total weight is estimated to amount to about 6,648 ton (1,575 ton of blanket module, 3,777 ton of back-plate, 372 ton of conducting shell and 924 ton of divertor cassette). In base case, main parameters of DEMO reactor are 8.2 m of major radius and 1.35 GW of fusion output. The lifetimes of blanket segment and divertor cassette are assumed to be 2.2 years and 0.6 year, respectively, 52,487 ton wastes is generated in plant life of 20 years. Therefore, there is a concern that a contamination controlled area for the radioactive waste may increase because much the waste is generated in every replacement. In this paper, management scenario is proposed to reduce the radioactive waste. The back-plates and cassette bodies (628 ton) of divertor was reused. As a result, the displacement per atom (DPA) of the back-plates of SUS316L was 0.2 DPA/year and that of the cassette bodies of F82H was 0.6 DPA/year. Therefore, reusing the back-plates and cassette bodies would be possible, if re-welding points are arranged under neutron shielding. It was found that radioactive waste could be reduced to 20%, when tritium breeding materials are recycled. Finally, a design of DEMO building such as a hot cell and temporary storage etc. is proposed.
Someya, Yoji; Tobita, Kenji; Tanigawa, Hisashi; Uto, Hiroyasu; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke
Proceedings of 23rd International Conference on Nuclear Engineering (ICONE-23) (DVD-ROM), 6 Pages, 2015/05
This paper presents neutronics analysis mainly focused on key design issues for self-sufficient tritium production based on the conceptual design study carried out for a fusion DEMO reactor in past several years, which includes new findings regarding design methodology of breeding blanket. Self-sufficient production of tritium is one of the most critical requirements for fusion reactors. We considered a fusion DEMO reactor with a major radius of about 8 m and fusion output of 1.5 GW with breeding blanket consisting of a mixed bed of LiTiO and BeTi pebbles. The net tritium breeding ratio (TBR) was estimated to be 1.15 with a three-dimensional analysis with the MCNP-5 with nuclear library of FENDL-2.1, satisfying a self-sufficient supply of tritium (net TBR
1.05). Throughout the research, we found that tritium breeding capability (i.e., local TBR) of breeding blanket is less dependent on the arrangement of cooling pipe in the blanket when the neutron wall loading is lower than about 1.5 MW/m
which is met in the DEMO considered. The result suggests that tolerance for the installation of cooling pipes in each blanket module will not be a critical matter. In addition, we found that a gap of about 0.02 m between neighboring blanket modules has little impact on the gross TBR.
Sakamoto, Yoshiteru; Nakamura, Makoto; Tobita, Kenji; Uto, Hiroyasu; Someya, Yoji; Hoshino, Kazuo; Asakura, Nobuyuki; Tokunaga, Shinsuke
Fusion Engineering and Design, 89(9-10), p.2440 - 2445, 2014/10
Times Cited Count：6 Percentile：42.97(Nuclear Science & Technology)Several concepts of DEMO have been proposed so far with plasma physics assumptions. At the same time, plasma performances foreseen in DEMO have been developed experimentally in tokamaks. However there are large gaps between the physics design parameters of the DEMO concepts and the simultaneous achieved parameters in tokamak experiments. Since one of the foreseeable integrated plasma performances is the ITER steady-state scenario, the projection of the scenario parameter to DEMO concept has been analyzed by using the systems code. The fusion power of 1GW can be obtained with the plasma major radius of 9 m. The same power can be obtained with 8 m if the distance between TF coil and plasma surface is reduced from 2 m to 1.5 m. Furthermore, it was found that the heat load on the divertor region is increased with increasing the normalized density and is decreased with increasing the normalized beta.
Someya, Yoji; Tobita, Kenji; Yanagihara, Satoshi*; Kondo, Masatoshi*; Uto, Hiroyasu; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Sakamoto, Yoshiteru
Fusion Engineering and Design, 89(9-10), p.2033 - 2037, 2014/10
Times Cited Count：9 Percentile：57.19(Nuclear Science & Technology)In the replacement period of a fusion power reactor, the assembly of blanket or divertor modules need to be removed from the reactor in order to minimize remote maintenance in the vacuum vessel and to attain a reasonable plant availability. In the hot cell, the modules will be removed from the backplate of the assembly. Here, note that the active cooling must be done by a way that does not cause contamination of the hot cell environment due to dispersion of tritium and tungsten dust. In this sense, the cooling scenario is adopted that the existing pipe of cooling water in the assembly is connected to a different cooling water system in the hot cell. In this scenario, the temperature of the assembly is maintained about 40-100C. On the other hand, the structural material (RAFM) of the blanket and divertor is not recycled due to its high contact dose rate. It should be crushed into small pieces to reduce volume of the waste and required storage space. Here, the decay heat must be removed by natural convection to keep the temperature below 65C for preventing water evaporation from the mortar. The RAFM is kept in the interim storage during 12 years until the required temperature conditions for mortar are ensured and then is disposed of.
Uto, Hiroyasu; Asakura, Nobuyuki; Tobita, Kenji; Sakamoto, Yoshiteru; Someya, Yoji; Hoshino, Kazuo; Nakamura, Makoto
Fusion Engineering and Design, 89(9-10), p.2456 - 2460, 2014/10
Times Cited Count：1 Percentile：8.88(Nuclear Science & Technology)Recently, use of an inter-linked (IL) superconducting coils in a tokamak fusion DEMO reactor were proposed. A basic idea of the IL-CS concept is to wind a CS such that it is linked in a set of toroidal field (TF) coils. In this presentation, the detailed descriptions of the engineering design of the superconducting CS linked in TFCs will be presented. Handling of a large exhausted power from the core plasma is the most important issue for the fusion reactor. Recently, advanced divertor concepts of super-X divertor (SXD) was proposed. The plasma equilibrium calculations for SlimCS showed that large coil currents are required for the conventional divertor coil location outside TFC. These results show that installation of the divertor coils inter-TFC (inter-linked PF) is required for the DEMO advanced divertor design. In this presentation, engineering feasibility of the inter-linked superconducting CS and PF for constructing the SXD equilibrium configuration will be presented.
Kolbasov, B. N.*; El-Guebaly, L.*; Khripunov, V. I.*; Someya, Yoji; Tobita, Kenji; Zucchetti, M.*
Fusion Engineering and Design, 89(9-10), p.2013 - 2017, 2014/10
Times Cited Count：9 Percentile：57.19(Nuclear Science & Technology)Within the framework of the International Energy Agency Program on Environmental, Safety and Economic Aspects of Fusion Power, an international collaborative study on management of fusion radioactive materials has been carried out to examine the back-end of the materials cycle. The strategy for handling fusion activated materials calls for three potential schemes: clearance, recycling and disposal. There is a growing international effort to avoid geologic disposal, for fusion in particular. Plasma facing components (divertor and blanket) normally contain high radioactivity and are not clearable. As clearance of sizeable components (such as biological shield, cryostat vessel, vacuum vessel, and some constituents of magnets) is highly desirable, we identified the source of radioisotopes that hinder the clearance of these components and investigated the impact of impurity control. Another study assessed radioactivity build up under repeated use of the divertor made of W-La
O
alloy. Effect of impurities on activated materials management is illustrated by the example of carbon-14 generation. Other studies examined impurities activation in concrete of biological shield and the impact of a specific activated materials scenario on the hot cell design and waste storage requirements.
Nakamura, Makoto; Tobita, Kenji; Gulden, W.*; Watanabe, Kazuhito*; Someya, Yoji; Tanigawa, Hisashi; Sakamoto, Yoshiteru; Araki, Takao*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Fusion Engineering and Design, 89(9-10), p.2028 - 2032, 2014/10
Times Cited Count：13 Percentile：70.2(Nuclear Science & Technology)After the Fukushima Dai-ichi nuclear accident, a social need for assuring safety of fusion energy has grown gradually in the Japanese (JA) fusion research community. DEMO safety research has been launched as a part of BA DEMO Design Activities (BA-DDA). This paper reports progress in the fusion DEMO safety research conducted under BA-DDA. Safety requirements and evaluation guidelines have been, first of all, established based on those established in the Japanese ITER site invitation activities. The amounts of radioactive source terms and energies that can mobilize such source terms have been assessed for a reference DEMO, in which the blanket technology is based on the Japanese fusion technology R&D programme. Reference event sequences expected in DEMO have been analyzed based on the master logic diagram and functional FMEA techniques. Accident initiators of particular importance in DEMO have been selected based on the event sequence analysis.
Nakamura, Makoto; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Sakamoto, Yoshiteru; Araki, Takao*; Watanabe, Kazuhito*; Matsumiya, Hisato*; Ishii, Kyoko*; et al.
Plasma and Fusion Research (Internet), 9, p.1405139_1 - 1405139_11, 2014/10
Key aspects of the safety study of a water-cooled fusion DEMO reactor is reported. Safety requirements, dose target, DEMO plant model and confinement strategy of the safety study are briefly introduced. The internal hazard of a water-cooled DEMO, i.e. radioactive inventories, stored energies that can mobilize these inventories and accident initiators and scenarios, are evaluated. It is pointed out that the enthalpy in the first wall/blanket cooling loops, the decay heat and the energy potentially released by the Be-steam chemical reaction are of special concern for the water-cooled DEMO. An ex-vessel loss-of-coolant of the first wall/blanket cooling loop is also quantitatively analyzed. The integrity of the building against the ex-VV LOCA is discussed.
Nakamura, Makoto; Ibano, Kenzo*; Tobita, Kenji; Someya, Yoji; Tanigawa, Hisashi; Gulden, W.*; Ogawa, Yuichi*
Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 8 Pages, 2014/10
Of late in Japan, a design study has been undertaken of a tokamak fusion DEMO with pressurized water coolant and solid pebble bed breeding blanket, but safety characteristics of this type of DEMO have not been well examined. In this paper, thermohydraulics analysis of in-vessel and ex-vessel loss-of-coolant accidents of a water-cooled tokamak DEMO is reported. Safety characteristics of water-cooled DEMO, particularly possible loads onto confinement barriers, are discussed based on the thermohydraulics analysis results. Measures to reduce such loads are also proposed.
Hoshino, Kazuo; Asakura, Nobuyuki; Shimizu, Katsuhiro; Tokunaga, Shinsuke; Takizuka, Tomonori*; Someya, Yoji; Nakamura, Makoto; Uto, Hiroyasu; Sakamoto, Yoshiteru; Tobita, Kenji
Plasma and Fusion Research (Internet), 9(Sp.2), p.3403070_1 - 3403070_8, 2014/06
no abstracts in English
Hirose, Takanori; Someya, Yoji; Tanigawa, Hisashi; Suzuki, Satoshi
Yosetsu Gakkai-Shi, 83(1), p.70 - 77, 2014/01
no abstracts in English
Uto, Hiroyasu; Someya, Yoji; Tobita, Kenji; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto
Nuclear Fusion, 53(12), p.123005_1 - 123005_8, 2013/11
Times Cited Count：10 Percentile：41.09(Physics, Fluids & Plasmas)Asakura, Nobuyuki; Shinya, Kichiro*; Tobita, Kenji; Hoshino, Kazuo; Shimizu, Katsuhiro; Uto, Hiroyasu; Someya, Yoji; Nakamura, Makoto; Ono, Noriyasu*; Kobayashi, Masahiro*; et al.
Fusion Science and Technology, 63(1T), p.70 - 75, 2013/05
no abstracts in English
Asakura, Nobuyuki; Shinya, Kichiro*; Tobita, Kenji; Hoshino, Kazuo; Shimizu, Katsuhiro; Uto, Hiroyasu; Someya, Yoji; Nakamura, Makoto; Ono, Noriyasu*; Kobayashi, Masahiro*; et al.
Fusion Science and Technology, 63(1T), p.70 - 75, 2013/05
Times Cited Count：14 Percentile：72.14(Nuclear Science & Technology)Design study of poloidal field coil (PFC) locations and current distribution for the advanced divertor in the Demo tokamak reactor was presented. Concept of the super-X divertor (SXD) for Demo reactor has an outer divertor leg longer than the conventional divertor, and it extends outboard to increase both the target wetted area and connection length to the outer target (
). Equilibrium calculation code, TOSCA, was developed by introducing two parameters, i.e. super-X null radius (
) and a ratio of the poloidal flux at the super-X null to that at the separatrix (
). Some SXD magnetic configurations with minimal number of PFCs located outside toroidal field coil (TFC) were Demonstrated. Locations of the divertor target were also investigated. It was found that the flux expansion can be increased up to 4-10 depending on the target location and , and that SXD has an advantage to increase with . Thus, the divertor plasma temperature is expected to decrease at the same upstream plasma density. On the other hand, large currents for the divertor PFCs were necessary. Other arrangements of PFCs such as (1) larger and (2) inside TFC, can reduce the PFC currents.
Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Hoshino, Kazuo; Asakura, Nobuyuki; Nakamura, Makoto; Tanigawa, Hisashi; Enoeda, Mikio; Tanigawa, Hiroyasu; Nakamichi, Masaru; et al.
Proceedings of 24th IAEA Fusion Energy Conference (FEC 2012) (CD-ROM), 8 Pages, 2013/03
This paper presents the conceptual design of a blanket with simplified structure whose interior consists of the mixture of breeder and multiplier pebble bed, cooling tubes and support for them only. Neutronics calculation indicated that the blanket satisfies a self-sufficient production of tritium. An important finding is that little decrease is seen in TBR even when the gap between neighboring blanket modules is as wide as 0.03 m. This means that blanket modules can be arranged with such a significant clearance gap without sacrifice of tritium production. On the other hand, the thickness of blanket housing is important from the viewpoint of safety. The blanket housing may rupture when the cooling pipe in the blanket is tearing, because thickness of structure materials is thin as 22 mm. This thickness is expected to maintain to 8 MPa in the steam pressure. Finally, the blanket housing, and aspect ratio of blanket shape is proposed in consideration of TBR, and engineering problem such as maintenance and manufacture are discussed.
