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

https://doi.org/10.1016/j.fusengdes.2014.04.062

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.

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

https://doi.org/10.1016/j.fusengdes.2012.03.024

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.

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

https://doi.org/10.1016/j.fusengdes.2011.01.141

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 LiSiO pebbles or LiO pebbles for the tritium breeding and BeTi 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 LiO pebbles blanket mixed with BeTi pebbles is the most effective and the TBR is greater than 1.05.

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

https://doi.org/10.1016/j.fusengdes.2011.03.022

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).

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

https://doi.org/10.1585/pfr.6.2405053

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. Several concepts on WCLL blanket and their engineering problems are presented.

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

Design concept of conducting shell for remote maintenance scheme in DEMO reactor

Uto, Hiroyasu; Takase, Haruhiko; Sakamoto, Yoshiteru; Tobita, Kenji; Hiwatari, Ryoji; Mori, Kazuo; Kudo, Tatsuya; Someya, Yoji; Asakura, Nobuyuki

no journal, , 

This presentation describes conceptual design of in-vessel component including conducting shell in DEMO design activities, in order to propose feasible DEMO reactor from plasma vertical stability and engineering viewpoint. The conducting shell design for plasma vertical stability was clarified from the plasma vertical stability analysis, and a conceptual design divided in the toroidal direction for the blanket remote maintenance was proposed. We evaluated dependence of the plasma vertical stability on the conducing shell parameters and electromagnetic force at plasma disruption by using a numerical simulation code (EDDYCAL) with actual shape and position of the vacuum vessel and in-vessel components. The calculation results showed that the double-loop shell has the most effect on plasma vertical stability. On the other hand, while 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.

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.

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

Safety studies for Japanese demo design with AINA code

Rivas, J. C.*; Nakamura, Makoto; Someya, Yoji; Takase, Haruhiko; Tobita, Kenji; de Blas, A.*; Dies, J.*; Fabbri, M.*; Riego, A.*

no journal, , 

Safety studies of plasma-wall transients have been performed with AINA code for the Japanese DEMO design (water cooled pebble bed). The AINA code has been adapted from its original mission of performing safety studies for ITER to this new mission. A breeding blanket model has been implemented in code. The configuration has been changed to implement the design parameters of DEMO reactor. First analyses performed show the behavior of the reactor during ex-vessel LOCA transients and during overpower events.