Helical reactor design FFHR-d1 and c1 for steady-state DEMO

Sagara, A.; Tamura, H.; Tanaka, Teruya; Yanagi, N.; Miyazawa, J.; Goto, Takuya; Sakamoto, R.; Yagi, Juro; Watanabe, Tsuguhiro; Takayama, S.

doi:10.1016/j.fusengdes.2014.02.076

Cited by 114 publications

(71 citation statements)

References 30 publications

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“…In the FFHR-d1, the blanket is divided into two components of tritium Breeding Blanket (BB) and neutron Shielding Blanket (SB). The molten salt FLiBe or FLiNaBe is chosen as the first option of the breeder material [1,2,7,8], mainly from the reason of safety. Selfcooling by the molten salt itself is considered as the first option and therefore no other coolant of water or helium gas is assumed for the BB in FFHR-d1.…”

Section: Author's E-mail: Miyazawa@lhdnifsacjpmentioning

confidence: 99%

“…• C [1]. The BB suffers from the direct irradiation of 14 MeV neutrons and therefore needs periodical replacement.…”

Section: Author's E-mail: Miyazawa@lhdnifsacjpmentioning

confidence: 99%

“…The conceptual design activity on a helical fusion reactor FFHR-d1 is ongoing [1,2]. The FFHR-d1 is basically a heliotron device similar to the LHD [3] and consists of two main helical coils and 4 planar poloidal field coils [4] of high-temperature superconducting (SC) magnet [5].…”

Section: Introductionmentioning

confidence: 99%

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Cartridge-Type Helical Blankets Aiming at Easy Construction and Maintenance for the FFHR-d1

Miyazawa

Tamura

Tanaka

et al. 2017

Plasma and Fusion Research

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A new cartridge-type blanket named the CARDISTRY-B is proposed for the helical fusion reactor FFHRd1. This blanket is composed of the neutron shield and the tritium breeder using molten salt. Both of these are toroidally segmented every two degrees. At each toroidal angle, the segmented parts are divided further into several cartridges in order to make it possible to assemble these cartridges after completion of the superconducting magnet coils. The neutron shield is basically assembled by using mortise and tenon prepared on each of the cartridges and the lower port, instead of the wide area welding. After assembly, the plasma side of the neutron shield is welded to form a vacuum vessel. Another side of the neutron shield facing on the superconducting magnet coils is covered with the thermal shield, which was already attached before assembly. The tritium breeder cartridges can be replaced without cutting or welding of cooling pipes inside the vacuum vessel, where severe radiation dose is expected. Details of the CARDISTRY-B, including the results of motion analysis for all cartridges and estimation of the cartridge weight, are discussed.

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Section: Author's E-mail: Miyazawa@lhdnifsacjpmentioning

confidence: 99%

“…• C [1]. The BB suffers from the direct irradiation of 14 MeV neutrons and therefore needs periodical replacement.…”

Section: Author's E-mail: Miyazawa@lhdnifsacjpmentioning

confidence: 99%

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Cartridge-Type Helical Blankets Aiming at Easy Construction and Maintenance for the FFHR-d1

Miyazawa

Tamura

Tanaka

et al. 2017

Plasma and Fusion Research

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“…On the basis of outputs from the LHD, design studies of the FFHR-d1 reactors have been performed. The FFHRd1 has a major radius of 15.6 m, a toroidal field of 4.7 T, and a fusion power of 3.0 GW [2,3]. Details of the threedimensional design of the superconducting magnets, consisting of two helical and four poloidal coils, and neutronics analyses have already been published [2].…”

Section: Introductionmentioning

confidence: 99%

“…The FFHRd1 has a major radius of 15.6 m, a toroidal field of 4.7 T, and a fusion power of 3.0 GW [2,3]. Details of the threedimensional design of the superconducting magnets, consisting of two helical and four poloidal coils, and neutronics analyses have already been published [2]. The up-todate neutronics analyses show that a maximum nuclear heat of 600 W/m 3 is generated in the helical coil [2].…”

Section: Introductionmentioning

confidence: 99%

A Cooling Concept for Indirectly Cooled Superconducting Magnets for the Fusion Reactor FFHR

Takahata

Tamura

Mito

et al. 2015

Plasma and Fusion Research

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Helical fusion power reactors have competitive advantages for steady-state operation because they use a currentless plasma. This paper presents the cooling concept of the indirectly cooled superconducting helical coil for the helical fusion reactor FFHR-d1. The helical coil consists of continuously wound aluminum-alloy-jacketed Nb 3 Sn superconductors and intermediate metal plates that not only cool the conductor indirectly, but also support the electromagnetic force. The copper cooling panels are partially mounted in the stainless steel intermediate plate and cooled by cryogenic supercritical helium. The conductor is cooled by contact with the cooling panels through an insulation material with high thermal conductivity, such as a ceramic. Fundamental calculations show that this cooling concept is technically feasible. Experimental investigations yielded a candidate for the ceramic insulator with high thermal conductivity.

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Nuclear Fusion

Yamada¹

2021

Handbook of Climate Change Mitigation and Adaptation

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Helical reactor design FFHR-d1 and c1 for steady-state DEMO

Cited by 114 publications

References 30 publications

Cartridge-Type Helical Blankets Aiming at Easy Construction and Maintenance for the FFHR-d1

Cartridge-Type Helical Blankets Aiming at Easy Construction and Maintenance for the FFHR-d1

A Cooling Concept for Indirectly Cooled Superconducting Magnets for the Fusion Reactor FFHR

Nuclear Fusion

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