Technological assessment between vertical and horizontal remote maintenance schemes for DEMO reactor

Utoh, Hiroyasu; Hiwatari, Ryoji; Tobita, K.; Someya, Youji; Aoki, Akira; Sakamoto, Y.; Asakura, N.

doi:10.1016/j.fusengdes.2017.03.036

Cited by 5 publications

(3 citation statements)

References 9 publications

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“…This technology is not yet proven, but if developed, would greatly facilitate the nuclear testing and breeding program, with the potential to change out walls and major components more rapidly. For instance, a vertical change out approach such as proposed in [52] may be attractive. With the high fields and compactness of this device, mechanical stresses from electromagnetic loads will be a critical issue, as maximum fields in the magnets, B max will reach levels ∼16-18 T for a 7 T device.…”

Section: Engineering Considerationsmentioning

confidence: 99%

The advanced tokamak path to a compact net electric fusion pilot plant

et al. 2021

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Physics-based simulations project a compact net electric fusion pilot plant with a nuclear testing mission is possible at modest scale based on the advanced tokamak concept, and identify key parameters for its optimization. These utilize a new integrated 1.5D core-edge approach for whole device modeling to predict performance by self-consistently applying transport, pedestal and current drive models to converge fully non-inductive stationary solutions, predicting profiles and energy confinement for a given density. This physics-based approach leads to new insights and understanding of reactor optimization. In particular, the levering role of high plasma density is identified, which raises fusion performance and self-driven ‘bootstrap currents’, to reduce current drive demands and enable high pressure with net electricity at a compact scale. Solutions at 6–7 T, ∼4 m radius and 200 MW net electricity are identified with margins and trade-offs possible between parameters. Current drive comes from neutral beam and ultra-high harmonic (helicon) fast wave, though other advanced approaches are not ruled out. The resulting low recirculating power in a double null configuration leads to a divertor heat flux challenge that is comparable to ITER, though reactor solutions may require more dissipation. Strong H-mode access (x2 margin over L–H transition scalings) and ITER-like heat fluxes are maintained with ∼20%–60% core radiation, though effects on confinement need further analysis. Neutron wall loadings appear tolerable. The approach would benefit from high temperature superconductors, as higher fields would increase performance margins while potential for demountability may facilitate nuclear testing. However, solutions are possible with conventional superconductors. An advanced load sharing and reactive bucking approach in the device centerpost region provides improved mechanical stress handling. The prospect of an affordable test device which could close the loop on net-electric production and conduct essential nuclear materials and breeding research is compelling, motivating research to validate the techniques and models employed here.

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Section: Engineering Considerationsmentioning

confidence: 99%

The advanced tokamak path to a compact net electric fusion pilot plant

et al. 2021

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“…In the JA DEMO, a banana segment and divertor cassette maintenance scheme were selected as a result of considerations of a variety of RM concepts. [39][40][41][42] The tentative foreseen replacement frequency of BB segments and divertor cassettes is about 4 to 5 years and 1 to 2 years, respectively. Part of divertor heat sink is composed of Cualloy cooling pipe with tungsten monoblock armor.…”

Section: Remote Maintenancementioning

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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“…Therefore, the divertor units are replaced together with the 140 BB cartridges. This number is equivalent to the total number of the components that need periodical maintenance in the tokamak DEMO of 128, which consists of 80 blanket segments and 48 divertor cassettes [10,11]. The other BB cartridges of BB14 -BB16 can be replaced if necessary.…”

Section: Overview Of the Cardistry-b2mentioning

confidence: 99%

Improved Design of a Cartridge-Type Helical Blanket System for the Helical Fusion Reactor FFHR-b1

Miyazawa

Tamura

Tanaka

et al. 2019

Plasma and Fusion Research

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The cartridge-type helical blanket system called the CARDISTRY-B has been proposed for the helical fusion reactor FFHR-d1. The CARDISTRY-B is aimed at easy construction and maintenance. However, there remain many issues in the design. For example, these include a large number of cartridges, no tangential ports, low compatibility with the helical divertor, and others. To solve these issues, an improved design of the cartridge-type blanket is proposed and named the CARDISTRY-B2. This is composed of the tritium breeding blanket (BB) filled with the flowing molten salt and the neutron shielding blanket (SB) filled with the tungsten-carbide. The number of the BB (or, SB) cartridges in the 1/10 section of the full torus are reduced from 32 to 17 (or, from 44 to 17). Tangential ports for the neutral beam injection heating are newly equipped together with the multipurpose blanket cartridge. Compatibility with the helical divertor is also improved. The BB cartridges are fixed on the fixing units set on the top of the SB. No other part of the BB is touching the SB except the fixing ribs. The SB cartridges surround the superconducting magnet coils and work as the cryostat inside the vacuum vessel. This makes the large bellows for the large maintenance ports unnecessary. Details of the CARDISTRY-B2 designed for a small helical fusion reactor FFHR-b1 are described in this paper.

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Technological assessment between vertical and horizontal remote maintenance schemes for DEMO reactor

Cited by 5 publications

References 9 publications

The advanced tokamak path to a compact net electric fusion pilot plant

The advanced tokamak path to a compact net electric fusion pilot plant

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

Improved Design of a Cartridge-Type Helical Blanket System for the Helical Fusion Reactor FFHR-b1

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