Research and development status on fusion DEMO reactor design under the Broader Approach

Tobita, K.; Federici, G.; Okano, Kunihiko

doi:10.1016/j.fusengdes.2014.02.077

Cited by 22 publications

(8 citation statements)

References 17 publications

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“…This is because the inter-link winding has difficulties in in-situ winding, the low alternating-current loss superconductor which is operative under high field, and the assembly process of TF coils, inter-linked PF coils and the vacuum vessel (VV). 19 The divertor simulation for SXD shows that a significant radiation region is produced between the super-X null (second null point) and the divertor target, and that the radiation loss in the divertor increases, producing fully detached plasma efficiently. 17 However, the resulting heat load to the divertor is critical in light of an engineering constraint in that a concentration of radiated power reaches 10 MW/m 2 on the divertor target plate.…”

Section: Ivb Divertor Configurationsmentioning

confidence: 99%

Japan’s Efforts to Develop the Concept of JA DEMO During the Past Decade

Tobita

Hiwatari

Sakamoto

et al. 2019

Fusion Science and Technology

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This paper summarizes the evolution of Japanese DEMO design studies in a retrospective manner by highlighting efforts to resolve critical design issues on DEMO. Japan is currently working on the conceptual study of a steady-state DEMO (JA DEMO) with a major radius Rp of 8.5 m and fusion power Pfus of 1.5 to 2 GW based on water-cooled solid breeding blanket with pressurized water reactor water condition (290ºC to 325ºC, 15.5 MPa). Such a lower Pfus allows to find realistic design solutions for divertor heat removal. Recognizing that divertor heat removal is one of the most challenging issues on DEMO, the divertor design has been carried out in different approaches, including numerical divertor plasma simulation, magnetic configurations, heat sink design, etc. It is noteworthy that the latest divertor simulation led to a design window allowing divertor heat removal of the peak heat flux of <10 MW/m 2 . The breeding blanket (BB) design has been concentrated on simplification of the internal structure and pressure tightness of the BB casing against the in-box loss-of-coolant accident. Due to a large amount of radioactive waste generated in periodic replacement of in-vessel components, downsizing of waste-related facilities has come to be regarded as a significant design issue. A possible waste management for reducing temporary waste storage was proposed, and its impact on the plant layout was assessed.

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Section: Ivb Divertor Configurationsmentioning

confidence: 99%

Japan’s Efforts to Develop the Concept of JA DEMO During the Past Decade

Tobita

Hiwatari

Sakamoto

et al. 2019

Fusion Science and Technology

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“…The reactor should be operated with a closed fuel-cycle and with realistic scenarios, and should be designed with feasible engineering solutions. The DDA has addressed design features of fusion DEMO reactors, critical design issues found through gap study between ITER and DEMO, and various efforts for resolving the issues [8,9]. The key achievements of the design activities can be summarized as below.…”

Section: Demo Design Activitymentioning

confidence: 99%

Recent technical progress on BA Program: DEMO activities and IFMIF/EVEDA

Yamanishi

Asakura

Tobita

et al. 2016

Fusion Engineering and Design

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“…In the DEMO design activity under the Broader Approach (BA) [1], Japan home team has proposed a water-cooled fusion DEMO reactor [2,3]. In this reactor, design of blanket is based on water cooled solid breeder.…”

Section: Introductionmentioning

confidence: 99%

Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

Watanabe

Nakamura

Tobita

et al. 2015

2015 IEEE 26th Symposium on Fusion Engineering (SOFE)

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Safety studies of a water-cooled fusion DEMO reactor have been performed. In the DEMO design, the blanket primary cooling system involves a large amount of energy due to pressurized water coolant (290-325 °C, 15.5 MPa). Moreover, it contains radioactive materials such as tritium and activated corrosion products. Therefore, 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 options of confinement strategies. In each option, the pressure and thermal loads to the confinement boundaries and total mass of tritium released to the environment were analyzed by accident analysis code MELCOR modified for fusion reactor. These analyses developed design parameters to maintain the integrity of the confinement boundaries. Keywords-water-cooled fusion DEMO, safety study, accident scenario analysis, safety system978-1-4799-8264-6/15/$31.00 ©2015 IEEE

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Research and development status on fusion DEMO reactor design under the Broader Approach

Cited by 22 publications

References 17 publications

Japan’s Efforts to Develop the Concept of JA DEMO During the Past Decade

Japan’s Efforts to Develop the Concept of JA DEMO During the Past Decade

Recent technical progress on BA Program: DEMO activities and IFMIF/EVEDA

Analysis on ex-vessel loss of coolant accident for a water-cooled fusion DEMO reactor

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