Fabrication of nitrogen trapping test loop for IFMIF-EVEDA

Yagi, Juro; Ban-nai, Mikiko; Suzuki, Akihiro; Terai, Takayuki; Nakamura, Kazuyuki; Kondo, Hiroo; Ida, Masato; Kanemura, Takuji; Furukawa, Tomohiro; Hirakawa, Yasushi; Anzai, Masaaki

doi:10.1016/j.fusengdes.2011.01.096

Cited by 9 publications

(4 citation statements)

References 7 publications

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“…NIFS and Universities have contributed significantly to the engineering validation of IFMIF for the Li target and the test cell systems. In particular, since the control of flow and the chemistry of Li is the commonly necessary technology to the liquid Li breeding blankets and divertors, infrastructure and experiences in Universities such as Li loop flow test 123 , deuterium 124 and nitrogen 125 showing that recycling and reuse after 10 years' would be possible by reducing Ti level below 1%, the alloy composition is being re-optimized by increasing Cr level for compensating mechanical property degradation by the decrease in Ti level below 1 %. Strict control of impurities which can produce long radioactivity by exposure to neutrons is also necessary.…”

Section: Design Of Laser Inertial Fusion Reactors and The Relevant Re...mentioning

confidence: 99%

Overview of Fusion Engineering Research in Japan Focusing on Activities in NIFS and Universities

Muroga

Fukada

Hayashi

2019

Fusion Science and Technology

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This paper provides an overview of Japanese fusion engineering research activities focusing on those being carried out by NIFS and Japanese universities (Universities). NIFS is promoting the Fusion Engineering Research Project as one of the three research projects. The majority of the activity in the project is being carried out by collaboration with Universities. Utilizing core facilities installed in NIFS and the unique infrastructures of Universities, collaboration between NIFS and Universities is performed in superconducting magnet, liquid breeder blanket, advanced materials, high heat flux components, and tritium safety. NIFS also carries out international collaboration programs, such as Japan-China, Japan-USA, and IEA-based collaborations, promoting participation of University researchers. Responsibility division with QST, and contributions to ITER-BA and "Action Plan toward DEMO Development" are also reported.

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Section: Design Of Laser Inertial Fusion Reactors and The Relevant Re...mentioning

confidence: 99%

Overview of Fusion Engineering Research in Japan Focusing on Activities in NIFS and Universities

Muroga

Fukada

Hayashi

2019

Fusion Science and Technology

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“…In the IFMIF, the cold traps (CTs) are used mainly to remove oxygen but not tritium since the T concentration is well below the T saturation limit of ~ 0.02% (by weight) at 200°C as noted below. Because of the very low T concentration, T was removed using a hot trap [35,37]. But in the present Li loop system, the T concentration in the LL is expected to be well above the T saturation limit, so that the recovery of T with the CT system is feasible [2].…”

Section: Tritium Removal By Cold Trapmentioning

confidence: 99%

Liquid lithium applications for solving challenging fusion reactor issues and NSTX-U contributions

Ono

Jaworski

Kaita

et al. 2017

Fusion Engineering and Design

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Steady-state fusion reactor operation presents major divertor technology challenges, including high divertor heat flux both steady-state and transients. In addition, there are unresolved issues of long term dust accumulation and associated tritium inventory and safety concerns [1]. It has been suggested that radiative liquid lithium divertor concepts with a modest lithium-loop could provide a possible solution for these outstanding fusion reactor technology issues, while potentially improving reactor plasma performance [2, 3]. The application of lithium (Li) in NSTX resulted in improved H-mode confinement, H-mode power threshold reduction, and reduction in the divertor peak heat flux while maintaining essentially Li-free core plasma operation even during H-modes. These promising results in NSTX and related modeling calculations motivated the radiative liquid lithium (LL) divertor (RLLD) concept [2] and its variant, the active liquid lithium divertor concept (ARLLD) [3], taking advantage of the enhanced non-coronal Li radiation in relatively poorly confined divertor plasmas. It was estimated that only a few moles/sec of lithium injection would be needed to significantly reduce the divertor heat flux in a tokamak fusion power plant. By operating at lower temperatures ≤ 450°C than the first wall ~ 600-700°C, the LL-covered divertor chamber wall surfaces can serve as an effective particle pump, as impurities generally migrate toward lower temperature LL divertor surfaces. To maintain the LL purity, a closed LL loop system with a modest circulating capacity of ~ 1 liter/second (l/sec) is envisioned to sustain the steady-state operation of a 1 GW-electric class fusion power plant. By running the Li loop continuously, it can carry the dust particles and impurities generated in the vacuum vessel to outside where the dust / impurities are removed by relatively simple filter and cold/hot trap systems. Using a cold trap system, it can recover tritium (T) in real time from LL at a rate of ~ 0.5 g / sec needed to sustain the fusion reaction while minimizing the T inventory With an expected T fraction of ≤ 0.7 %, an acceptable T inventory level can be achieved. In NSTX-U [4, 5], preparations are now underway to elucidate the physics of Li plasma interactions with a number of Li application tools and Li radiation spectroscopic instruments. The NSTX-U Li evaporator, which provides Li coating over the lower divertor plate, can offer important information on the RLLD concept, and the Li granule injector will test some of the key physics issues for the ARLLD concept. A LL loop is also being prepared off-line for prototyping future use on NSTX-U.

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“…The principle of the trapping is to use the absorption of the nitrogen into the Ti at higher temperature. Preliminary experiments have been done with rotating Fe-Ti disk, Fe-Ti alloy pebble and static Fe-Ti plate [4]. Fig.…”

Section: Nitrogen Hot Trapmentioning

confidence: 99%