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Journal Articles

Progress report of Japanese simulation research projects using the high-performance computer system Helios in the International Fusion Energy Research Centre

Ishizawa, Akihiro*; Idomura, Yasuhiro; Imadera, Kenji*; Kasuya, Naohiro*; Kanno, Ryutaro*; Satake, Shinsuke*; Tatsuno, Tomoya*; Nakata, Motoki*; Nunami, Masanori*; Maeyama, Shinya*; et al.

Purazuma, Kaku Yugo Gakkai-Shi, 92(3), p.157 - 210, 2016/03

The high-performance computer system Helios which is located at The Computational Simulation Centre (CSC) in The International Fusion Energy Research Centre (IFERC) started its operation in January 2012 under the Broader Approach (BA) agreement between Japan and the EU. The Helios system has been used for magnetised fusion related simulation studies in the EU and Japan and has kept high average usage rate. As a result, the Helios system has contributed to many research products in a wide range of research areas from core plasma physics to reactor material and reactor engineering. This project review gives a short catalogue of domestic simulation research projects. First, we outline the IFERC-CSC project. After that, shown are objectives of the research projects, numerical schemes used in simulation codes, obtained results and necessary computations in future.

Journal Articles

Design concept of conducting shell and in-vessel components suitable for plasma vertical stability and remote maintenance scheme in DEMO reactor

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:7 Percentile:17.68(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.

Journal Articles

Design study of blanket structure based on a water-cooled solid breeder for DEMO

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:20 Percentile:3.62(Nuclear Science & Technology)

Blanket concept with simplified interior for mass production has been developed with a mixed bed of Li$$_{2}$$TiO$$_{3}$$ and Be$$_{12}$$Ti pebbles, a coolant condition of 15.5 MPa and 290-325$$^{circ}$$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 300$$^{circ}$$C) 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.

Journal Articles

Management strategy for radioactive waste in the fusion DEMO reactor

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:8 Percentile:22.06(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.

Journal Articles

Simulation study of power load with impurity seeding in advanced divertor "short super-X divertor" for a tokamak reactor

Asakura, Nobuyuki; Hoshino, Kazuo; Shimizu, Katsuhiro; Shinya, Kichiro*; Uto, Hiroyasu; Tokunaga, Shinsuke; Tobita, Kenji; Ono, Noriyasu*

Journal of Nuclear Materials, 463, p.1238 - 1242, 2015/08

 Times Cited Count:10 Percentile:15.29(Materials Science, Multidisciplinary)

Arrangements of interlink divertor coils and divertor geometries for short super-X was proposed as the Demo advanced divertor design. Performance of plasma detachment under the large heat flux was investigated to optimize the divertor design, using SONIC simulation with Ar impurity seeding, where Pout = 500 MW, ne = 7$$times$$10$$^{19}$$ m$$^{-3}$$ at the core-edge boundary and the same diffusion coefficients for ITER simulation. Effects on the plasma temperature and density distributions were compared to the conventional divertor. The first run results with the same radiation power fraction of 0.92 in the conventional divertor showed that full detached plasma is produced, the maximum radiation region was maintained upstream the divertor target, and both the plasma heat load plus radiation load at the target was reduced to 10 MWmm$$^{-2}$$ level. Simulation for the lower radiation power fractions of 0.8-0.9 was also performed, and physics issues of the short super-X divertor are discussed.

Journal Articles

Neutronics analysis for fusion DEMO reactor design

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 Li$$_{2}$$TiO$$_{3}$$ and Be$$_{12}$$Ti 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$$^{2}$$ 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.

Journal Articles

Relationship between net electric power and radial build of DEMO based on ITER steady-state scenario parameters

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:4 Percentile:54.43(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.

Journal Articles

Studies of impurity seeding and divertor power handling in fusion reactor

Hoshino, Kazuo; Asakura, Nobuyuki; Shimizu, Katsuhiro; Tokunaga, Shinsuke

Proceedings of 25th IAEA Fusion Energy Conference (FEC 2014) (CD-ROM), 6 Pages, 2014/10

Journal Articles

Divertor study on DEMO reactor

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

Journal Articles

A Simulation study of large power handling in the divertor for a Demo reactor

Asakura, Nobuyuki; Shimizu, Katsuhiro; Hoshino, Kazuo; Tobita, Kenji; Tokunaga, Shinsuke; Takizuka, Tomonori*

Nuclear Fusion, 53(12), p.123013_1 - 123013_15, 2013/12

 Times Cited Count:31 Percentile:8.85(Physics, Fluids & Plasmas)

Power exhaust scenario for 3 GW fusion reactor has been developed with enhancing the radiation loss from seeding impurity. Impurity transport and plasma detachment were simulated self-consistently under the Demo divertor condition using an integrated divertor code SONIC. Power handling with different seeding impurities showed that the total heat load, including the plasma transport and radiation, was reduced from 15 MW/m$$^{2}$$ to 10 MW/m$$^{2}$$ for the higher Z case, where both heat load components became comparable. Effects of a long-leg divertor were examined. Full detachment was obtained and peak heat load was decreased to 12 MW/m$$^{2}$$, where neutral load became comparable to the plasma and radiation load, which may be due to small flux expansion. Finally, plasma diffusion significantly affected the performance of the plasma detachment. Peak heat lard was reduced to 5 MW/m$$^{2}$$ well below engineering design level, with uniformly increasing the diffusion coefficient by the factor of two.

Oral presentation

Conceptual design of inter-linked superconducting coils for advanced divertor configuration of BA DEMO reactor

Uto, Hiroyasu; Asakura, Nobuyuki; Tobita, Kenji; Someya, Yoji; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke

no journal, , 

Recently, use of an inter-linked (IL) superconducting coils in a tokamak fusion DEMO reactor were proposed. A basic idea of the IL-central solenoid (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 TFC 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 (IL-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.

Oral presentation

Investigation of physics and engineering issues in advanced divertor design for tokamak Demo reactor

Asakura, Nobuyuki; Shinya, Kichiro*; Hoshino, Kazuo; Uto, Hiroyasu; Someya, Yoji; Tokunaga, Shinsuke; Tobita, Kenji; Nakamura, Makoto; Sakamoto, Yoshiteru; Ono, Noriyasu*

no journal, , 

no abstracts in English

Oral presentation

Conceptual study of maintenance for fusion DMEO reactor toward reduction of radioactive waste

Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Sakamoto, Yoshiteru; Tokunaga, Shinsuke

no journal, , 

We are considering the reduction of radioactive waste which is generated in every replacement of an in-vessel component such as a blanket segment and divertor cassette for fusion DEMO reactor. Main parameters of DEMO reactor are 8.2 m of major radius and 1.35 GW of fusion output. 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 5,630 ton (1,930 ton of blanket module, 3,700 ton of back-plate and 990 ton of divertor cassette). The lifetimes of blanket segment and divertor cassette are assumed to be 3 years and 1 year, respectively, 8,600 ton wastes is generated in 3 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 of three dimensional neutron transportation code MCNP, the displacement per atom (DPA) of the back-plates of SUS316L was about 0.6 DPA/3years and that of the cassette bodies of F82H was 1.0 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 in 3 years could be reduced by 68%.

Oral presentation

Study of design region to resolve DEMO issues in BA activity

Sakamoto, Yoshiteru; Nakamura, Makoto; Tobita, Kenji; Uto, Hiroyasu; Someya, Yoji; Hoshino, Kazuo; Asakura, Nobuyuki; Tokunaga, Shinsuke

no journal, , 

Relations between a net electrical output power and dimensions of components in radial build are investigated based on the ITER plasma performance to develop a conceptual design of DEMO with the net electrical output power of several hundred MW. Reducing the dimensions of in-vessel components and increasing the thickness of the toroidal field coil contribute to strengthen the toroidal magnetic field at plasma, which brings about increase in a net electrical output power. The relation between the minimum plasma major radius and the maximum net electrical output power is clarified. Furthermore effects of improvements in the ITER plasma performance on the net electricity are also analyzed; indicating the increase of normalized beta could have advantage from the viewpoint of the divertor heat load because the increase of synchrotron radiation loss power contributes to reduce the divertor heat load, though the higher energy confinement is required.

Oral presentation

Integrated simulation study of fusion output control in DEMO

Tokunaga, Shinsuke; Sakamoto, Yoshiteru; Tobita, Kenji; Asakura, Nobuyuki; Hoshino, Kazuo; Someya, Yoji; Uto, Hiroyasu; Nakamura, Makoto

no journal, , 

no abstracts in English

Oral presentation

Investigation of advanced divertor, short super-X divertor, for demo tokamak reactor

Asakura, Nobuyuki; Hoshino, Kazuo; Uto, Hiroyasu; Someya, Yoji; Shimizu, Katsuhiro; Shinya, Kichiro*; Tokunaga, Shinsuke; Tobita, Kenji; Ono, Noriyasu*

no journal, , 

no abstracts in English

Oral presentation

Management strategy of radioactive waste in the fusion DEMO reactor

Someya, Yoji; Tobita, Kenji; Kondo, Masatoshi*; Yanagihara, Satoshi*; Uto, Hiroyasu; Sakamoto, Yoshiteru; Asakura, Nobuyuki; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke

no journal, , 

We are considering the reduction of radioactive waste which is generated in every replacement of an in-vessel component such as a blanket segment and divertor cassette for fusion DEMO reactor. Main parameters of DEMO reactor are 8.2 m of major radius and 1.35 GW of fusion output. 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. A conducting shell for plasma positional stability and high beta access is installed in between the blanket modules and 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). The lifetimes of blanket segment and divertor cassette are assumed to be 4 years and 1 year, respectively, 9,420 ton wastes is generated in 4 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 a hot cell, the blanket module is removed from the back-plate by a remote handing. A tritium, radioactive dust and decay heat are inherent in the in-vessel component. The hot cell environment for the maintenance is cooled by natural air convection to prevent diffusion of the tritium and radioactive dust. Since the hot cell made a concrete, on the other hand, the temperature of the hot cell must be maintained below 65$$^{circ}$$C for preventing water evaporation. As a result of thermal analysis, the maintenance of blanket segment is started after 6 years of storage until the temperature required for the concrete in the hot cell. 6 years after shutdown, a dominant contact dose rate of the component was 100 Sv/hour.

Oral presentation

Assessment of remote maintenance schemes for BA DEMO reactor

Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Kakudate, Satoshi; Tanigawa, Hisashi; Asakura, Nobuyuki; Sakamoto, Yoshiteru; Hoshino, Kazuo; Nakamura, Makoto; Tokunaga, Shinsuke

no journal, , 

Maintenance schemes are one of the critical issues in DEMO design, significantly affecting the configuration of in-vessel components, the size of toroidal field (TF) coil, the arrangement of poloidal field (PF) 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. The design study clarifies some assessment factors on DEMO remote maintenance scheme, for example, (1) minimization of the size and magnetic stored energy of TF coil and PF coils, (2) divertor maintenance. This presentation describes engineering design of each maintenance schemes and evaluation results of comparison among maintenance schemes.

Oral presentation

Physics and engineering studies of the advanced divertor for a fusion reactor

Asakura, Nobuyuki; Hoshino, Kazuo; Uto, Hiroyasu; Shinya, Kichiro*; Tokunaga, Shinsuke; Shimizu, Katsuhiro; Someya, Yoji; Tobita, Kenji; Ono, Noriyasu*

no journal, , 

A short super-X divertor is proposed as a new option for fusion divertor, where field line length from the divertor null to the outer target (L//div) was largely increased (more than two times) in a similar size of conventional divertor. Physics and engineering design studies have progressed. Minimal number of the divertor coils (1 or 2) were installed inside the toroidal field coil (TFC), i.e. interlink-winding (interlink). Arrangement of the poloidal field coils (PFCs) and their currents were determined, taking into account of the engineering design such as vacuum vessel and the neutron shield structures, and maintenance scenario of the divertor and blankets. Divertor plasma simulation showed that large radiation region is produced between the super-X null and the target, and the plasma temperature becomes low (1-2 eV) both at the inner and outer divertors, i.e. fully detached plasma was obtained efficiently.

Oral presentation

Conceptual design study of an advenaced divertor, short super-X divertor

Asakura, Nobuyuki; Hoshino, Kazuo; Uto, Hiroyasu; Shinya, Kichiro*; Shimizu, Katsuhiro; Tokunaga, Shinsuke; Someya, Yoji; Tobita, Kenji; Ono, Noriyasu*

no journal, , 

no abstracts in English

27 (Records 1-20 displayed on this page)