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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:8 Percentile:66.89(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

Study of safety features and accident scenarios in a fusion DEMO reactor

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:10 Percentile:68.92(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.

Journal Articles

Key aspects of the safety study of a water-cooled fusion DEMO reactor

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.

Journal Articles

Conceptual study of vertical sector transport maintenance for DEMO fusion reactor

Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Takase, Haruhiko

Fusion Engineering and Design, 87(7-8), p.1409 - 1413, 2012/08

 Times Cited Count:9 Percentile:62.49(Nuclear Science & Technology)

In BA DEMO design activity assessment of various maintenance schemes for DEMO reactor has been studied. The maintenance scheme is one of the critical issues for DEMO design, and required high availability. SlimCS designed in JAEA adopts the horizontal sector transport hot cell maintenance scheme. In order to decide a most probable DEMO reactor maintenance scheme, assessment of various maintenance schemes for DEMO are important. In this presentation the maintenance concept vertical sector transport is presented. In the sector maintenance scheme, the number of cutting/re-welding points of piping is minimized. The sector including blanket modules and high temperature shield was divided into 36 segments in toroidal direction. The sectors are removed and inserted through upper alternately-layered vertical maintenance ports. In the case of the vertical sector transport maintenance scheme, the inter-coil structures against turnover force in TF coils could be adopted.

Journal Articles

Nuclear analysis of DEMO water-cooled blanket based on sub-critical water condition

Liu, C.; Tobita, Kenji; Uto, Hiroyasu; Someya, Yoji; Takase, Haruhiko; Asakura, Nobuyuki

Fusion Engineering and Design, 86(12), p.2839 - 2842, 2011/12

 Times Cited Count:9 Percentile:62.21(Nuclear Science & Technology)

Journal Articles

Simplification of blanket system for SlimCS fusion DEMO reactor

Someya, Yoji; Takase, Haruhiko; Uto, Hiroyasu; Tobita, Kenji; Liu, C.; Asakura, Nobuyuki

Fusion Engineering and Design, 86(9-11), p.2269 - 2272, 2011/11

 Times Cited Count:27 Percentile:90.93(Nuclear Science & Technology)

Conceptual design of a tritium-breeding blanket for SlimCS has been studied. The blanket structure with neutron multiplier Be-plate was designed to be as thin as possible with keeping high Tritium Breeding Ratio (TBR). However, a structure of the blanket is complexity and the manufacture of the blanket is difficult from the viewpoint of engineering. Therefore, simplification of blanket structure is necessary for SlimCS. In this paper, we propose a simple blanket structure without decreasing the net TBR below 1.05. The proposed blanket structure is mixed Li$$_{4}$$SiO$$_{4}$$ pebbles or Li$$_{2}$$O pebbles for the tritium breeding and Be$$_{12}$$Ti pebbles for the neutron multiplication and these pebbles are filled in the blanket. ANIHEAT code with the nuclear data library FENDL-2.0 was used for the calculations of the neutronics and thermal analyses. As a result, it is shown that Li$$_{2}$$O pebbles blanket mixed with Be$$_{12}$$Ti pebbles is the most effective and the TBR is greater than 1.05.

Journal Articles

Development of a two-dimensional nuclear-thermal-coupled analysis code for conceptual blanket design of fusion reactors

Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Sato, Satoshi; Seki, Yohji; Takase, Haruhiko

Fusion Engineering and Design, 86(9-11), p.2378 - 2381, 2011/10

 Times Cited Count:10 Percentile:65.84(Nuclear Science & Technology)

For DEMO reactor blanket design, a two-dimensional (2-D) nuclear-thermal-coupled analysis code, DOHEAT, has been developed. In DOHEAT, the neutron flux is calculated by a 2-D transport code, DOT3.5, with the nuclear data library, FUSION-40, and the nuclear heating rate and the local TBR profile of blanket are calculated using the 2-D neutronics calculation code, APPLE-3. Use of the code has showed outstanding usefulness in the blanket design where detailed evaluation of neutron flux, nuclear heating rate, tritium breeding ratio (TBR) and the temperature of materials is required for various blanket concepts and trial-and-error-basis iteration is sometimes necessary. DOHEAT can replace the actual blanket structure by a more realistic model including cooling tubes, multipliers and breeders. A validation calculation indicates that DOHEAT provides reasonable results on the temperature profile.

Journal Articles

Maintenance concept for the SlimCS DEMO reactor

Tobita, Kenji; Uto, Hiroyasu; Kakudate, Satoshi; Takase, Haruhiko; Asakura, Nobuyuki; Someya, Yoji; Liu, C.

Fusion Engineering and Design, 86(9-11), p.2730 - 2734, 2011/10

 Times Cited Count:13 Percentile:74.14(Nuclear Science & Technology)

For high availability of DEMO operation, sector horizontal transport hot cell maintenance scheme was studied. Transport of sector with 730 tons is carried out using a wheeled platform. The driving force of pulling the sector into a cask is ball screws. The fulcrum of the ball screws is the cryostat wall so that a large pulling force is expected with no-counter balance. The cask containing the sector is delivered by air casters from the cryostat to the hot cell. For the maintenance scheme, new concepts such as transfer of the tilting forces of toroidal coils using ropes and shafts and supports for the tilting force using reinforced concrete floor or cryostat wall were proposed. Based on the maintenance concept, the period required for replacement of all sectors is estimated to be 35.5-67.5 days, satisfying the design target (shorter than 3 months).

Journal Articles

Blanket concept of water-cooled lithium lead with beryllium for the SlimCS fusion DEMO reactor

Uto, Hiroyasu; Tobita, Kenji; Someya, Yoji; Takase, Haruhiko; Asakura, Nobuyuki

Plasma and Fusion Research (Internet), 6, p.2405053_1 - 2405053_4, 2011/08

In this study, as an advanced option for SlimCS blanket, conceptual design study of water-cooled lithium lead (WCLL) blanket was performed. Compared with solid breeder, liquid lithium-lead (LiPb) breeder seems to have advantages of the sustainment of a design value of TBR independent of lithium burn-up and of a reduction of radioactive waste. However, in SlimCS, the net TBR supplied from WCLL blanket is not enough because the thickness of blanket in SlimCS is limited to 45 cm by conducting shell position for high beta access. Therefore, the beryllium (Be) pebble bed was adopted as additional multiplier. Considering of temperature of blanket materials, a double pipe structure was adopted. The Be pebble was separated by SiC/SiC composite tube, and was cooled by coolant on center. The local TBR of WCLL with Be blanket was similar to that of solid breeder blanket on the neutron wall load Pn = 5 MW/m$$^{2}$$. Several concepts on WCLL blanket and their engineering problems are presented.

Journal Articles

Comparison of coolant conditions in the blanket for a water-cooled DEMO reactor SlimCS

Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Takase, Haruhiko; Liu, C.; Asakura, Nobuyuki

Plasma and Fusion Research (Internet), 6, p.2405108_1 - 2405108_4, 2011/08

Conceptual design of an alternative tritium-breeding blanket for SlimCS has been studied. The proposed blanket concept is that Li$$_{4}$$SiO$$_{4}$$ pebbles or Li$$_{2}$$O pebbles for the tritium breeding and Be$$_{12}$$Ti pebbles for the neutron multiplication are mixed and these pebbles are filled in the blanket. The coolant condition was selected to be sub-critical water, whose temperature difference between inlet and outlet were 290$$^{circ}$$C and 360$$^{circ}$$C, respectively, and pressure was 23 MPa. When Li$$_{2}$$O pebbles were mixed with Be$$_{12}$$Ti pebbles, higher TBR was obtained, being greater than 1.05 for the blanket with the thickness of 0.48 m. However, the compatibility of the blanket structural material (F82H) with the sub-critical water is a concern. As the second step, therefore, we replaced the condition by the PWR water condition of 15.5 MPa and 290-330$$^{circ}$$C to improve the compatibility with F82H. In addition, the PWR water has an advantage that matured technologies in nuclear power plants will be likely to reduce development risks in fusion plant engineering. Therefore, consideration of coolant plumbing was decreased from all length in blanket. On the other hand, use of the PWR water to the blanket requires a reduction of coolant plumbing length to meet the temperature range. The proposed blanket was assessed with an ANIHEAT code, and the two cases of coolant conditions were compared.

Journal Articles

Comparison of tokamak DEMO reactor concepts between Japan and Europe and present status of the United States

Takase, Haruhiko

Denki Gakkai Kenkyukai Shiryo, Genshiryoku Kenkyukai (NE-10-002), p.5 - 8, 2010/08

Tokamak DEMO reactor design concepts of Japan and Europe are presented and present status of reactor design activities in the United States is introduced in this paper. The design concept by Europe has a priority in the early realization of fusion energy while the development strategy by the United States has a priority in the economy. On the other hand, the design concept by Japan has an intermediate characteristic between both strategies.

JAEA Reports

Pellet injection and plasma behavior simulation code PEPSI

Takase, Haruhiko*; Tobita, Kenji; Nishio, Satoshi

JAERI-Data/Code 2003-013, 46 Pages, 2003/08

JAERI-Data-Code-2003-013.pdf:1.59MB

no abstracts in English

Journal Articles

3-D electromagnetic transient characteristics of in-vessel components in tokamak reactor

Takase, Haruhiko; Senda, Ikuo; Araki, Masanori; Shoji, Teruaki; Tsunematsu, Toshihide

IAEA-CN-69/FTP/28, 4 Pages, 1998/00

no abstracts in English

Oral presentation

Development of a two-dimensional nuclear-thermal-coupled analysis code DOHEAT for DEMO reactor blanket design study

Uto, Hiroyasu; Tobita, Kenji; Sato, Satoshi; Seki, Yohji; Someya, Yoji; Takase, Haruhiko

no journal, , 

no abstracts in English

Oral presentation

The Design optimization of DEMO water-cooled blanket

Liu, C.; Tobita, Kenji; Uto, Hiroyasu; Takase, Haruhiko; Someya, Yoji; Asakura, Nobuyuki

no journal, , 

Oral presentation

Study of tokamak DEMO reactor designs by Japan and Europe

Takase, Haruhiko

no journal, , 

Tokamak DEMO reactor design concepts of Japan and Europe are presented and present status of reactor design activities in the United States is introduced in this paper. The design concept by Europe has a priority in the early realization of fusion energy while the development strategy by the United States has a priority in the economy. On the other hand, the design concept by Japan has an intermediate characteristic between both strategies.

Oral presentation

Burn fraction in a fusion reactor

Takase, Haruhiko

no journal, , 

no abstracts in English

Oral presentation

Study of operation area on DEMO reactor by system code

Uto, Hiroyasu; Tobita, Kenji; Takase, Haruhiko; Someya, Yoji; Asakura, Nobuyuki

no journal, , 

no abstracts in English

Oral presentation

Simplification of breeding blanket for fusion DEMO reactor

Someya, Yoji; Tobita, Kenji; Uto, Hiroyasu; Takase, Haruhiko; Liu, C.; Asakura, Nobuyuki

no journal, , 

no abstracts in English

Oral presentation

Analysis of plasma position control for Broader Approach (BA) DEMO reactor

Takase, Haruhiko; Tobita, Kenji; Sakamoto, Yoshiteru; Uto, Hiroyasu; Mori, Kazuo; Kudo, Tatsuya

no journal, , 

Analysis of plasma position control is one of important issues for design of DEMO reactor on Broader Approach (BA). Especially, plasma performance, blanket design and maintenance scheme influence the plasma position control mutually. Therefore, we made a numerical simulation code that consists of plasma equilibrium analysis, eddy current analysis and plasma motion analysis. Since we analyzed several cases of design using this numerical simulation code, the results will be shown.

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