Electrical insulation systems for the ITER CS modules

The Superconducting Conductor Testing Facility, which developed to evaluate the reliability of engineering technology and safe operation in fusion reactor operation environment., is under engineering design by ASIPP. In case of an emergency shut-down like in succession of a quench, the voltage across the coil may rise to about 2.5 kV. Therefore, the facility has to be provided with a reliable insulation. This paper introduces the preliminary design of the insulation system. The insulation system of Superconducting Conductor Testing Facility includes turn and ground insulation. Since the magnet of the facility adopts the manufacturing process of "Wind&React ", so its turn insulation material needs to withstand a heat treatment temperature of 650℃. The insulation system is composed of S-glass ber reinforced tape and vacuum-pressure impregnated in a DGEBF epoxy system. The mechanical properties are being subjected to investigations with respect to the design requirements and operating conditions.The test results are analyzed to verify the feasibility of insulation system design.

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“…( 6) Right low eld coil (LFC-2). (7) Connecting wire between medium eld coils. (8) Connecting wire between low eld coils.…”

Section: A Magnet Systemmentioning

confidence: 99%

Preliminary Design of Insulation System for Superconducting Conductor Testing Facility

Liu

Jin

et al. 2023

J Fusion Energ

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“…High current superconducting cables are typically needed to reduce coil inductance and allow quench protection in large scale insulated fusion magnets. This reduces the discharge voltage in coil insulation (i.e., limited to be within 20-30 kV during normal and fault conditions, for example as defined in ITER electrical design criteria [31][32]). For the cable-in-conduit (CICC) approach, it is unrealistic to model detailed performance and critical current behavior of each tape used in a HTS cable (stacked tape or others).…”

Section: Toroidal Field Coil Systemmentioning

confidence: 99%

Low cost, simpler HTS cable conductors for fusion energy systems

Zhai

Otto

Zarnstorff

2022

IOP Conf. Ser.: Mater. Sci. Eng.

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The cost and complexity of large, high-field superconducting magnet modules and related subsystems comprise 30% to 60% of the fusion reactor core capital cost. The strategic plan for the U.S. burning plasma research, the Fusion Energy Sciences Committee Report (FESAC) “Power the Future: Fusion and Plasmas’', and 2021 NASEM report “Key Goals and Innovations needed for a U.S. Fusion Pilot Plant” recommends that the U.S. pursue innovative science and technology to enable construction of a Fusion Pilot Plant (FPP) that produces net electricity from fusion at reduced capital cost. To achieve this, a novel combination of lower-cost high temperature superconductors (HTS) in cable configurations with co-wound reinforcement for higher current density are being investigated using a simplified construction strategy to produce compact stable coils. They would be capable of generating 20 T at up to 10-20 K. Small-scale, inexpensive test coils and prototypes will help develop each feature and validate cabled conductor design models. The near term goal is to validate engineering approaches, scientific models and fabrication capabilities applicable to fusion reactor development such as U.S. fusion nuclear science facility (FNSF), sustained high-power density tokamak facility (SHPD) and FPP designs. The design options include lower-cost, high-strength, quench resistant REBCO or Bi-2212 cables in an all metal coil design that simplifies HTS coil construction and quench protection system, with co-wound reinforcements that integrate stress management in HTS cable design and provides thermal mass to help prevent quench damage.

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“…The conventional organic insulation has an order of magnitude lower elastic modulus as compared to a jacketed cable-in-conduit (CIC) conductor. This difference made coils experience a spongy effect, as it creates a winding elastically soft and the sponginess increases the mechanical strain effect, in particular the insulation normal tensile strain when the coil is energized [12,13]. Conventional insulation composites also reduce the overall winding current density, while it electrically isolates turns in the winding and prevents current bypassing through adjacent turns during quench for coil protection.…”

Section: Introductionmentioning

confidence: 99%

Design, construction, and testing of no-insulation small subscale solenoids for compact tokamaks

Zhai

Berlinger

Barth

et al. 2021

Supercond. Sci. Technol.

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Fusion energy systems studies (FESS) for next-step devices based on the most promising magnetic configurations indicate that high magnetic fields and high current density for magnet coil systems may reduce device size and lower the cost. High current density and radiation resistant fusion magnets are particularly beneficial for low cost, low aspect ratio compact reactor designs such as Fusion Nuclear Science Facility (FNSF), Fusion Pilot Plant (FPP) of low-aspect ratio spherical tokamak (ST) or compact stellarators. Unlike typical high field magnets for nuclear magnetic resonance (NMR), magnetic resonance imaging (MRI), or high energy physics research applications, neutron irradiation damage to organic insulations in the coil winding pack is a critical issue for next generation fusion reactors where orders of magnitude higher neutron fluence than those in present experimental reactors such as ITER are expected. Moreover, a high coil winding pack current density is needed for high field magnets in low cost, compact radial build next step fusion reactors. The slow current charging time, however, is an issue in fully non-insulated coils. We present the design, construction and testing of subscale Nb3Sn solenoids, up to half the diameter of the central solenoid (CS) for National Spherical Torus Experiment (NSTX), with and without inter-layer insulation for enhancing radiation resistance while improving the winding pack current density and coil performance. The major radius for the NSTX/NSTX-Upgrade (reference for compact ST) is 0.854/0.934 meter, and 4.8 meters for FNSF. The magnetic field for NSTX/NSTX-U at the plasma center is 0.6/1 Tesla and 9 Tesla for FNSF. The inner diameter of the central solenoid coil is 0.2/0.4 meter for NSTX/NSTX-U and 1.2 meters for FNSF. The current charging and discharging behavior of the prototype solenoids was investigated to quantify performance advantages of a simplified coil fabrication without error-prone vacuum pressure impregnation (VPI) for the removal of epoxy organic insulation. Scalability of the no insulation concept for fusion can be addressed via engineered coating for novel insulation on contact resistance or intra-layer no insulations. Radiation resistance and coil winding efficiency were significantly improved in the no insulation coils for next step compact fusion reactors.

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Electrical insulation systems for the ITER CS modules

Cited by 10 publications

References 8 publications

Preliminary Design of Insulation System for Superconducting Conductor Testing Facility

Preliminary Design of Insulation System for Superconducting Conductor Testing Facility

Low cost, simpler HTS cable conductors for fusion energy systems

Design, construction, and testing of no-insulation small subscale solenoids for compact tokamaks

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