Hydraulic characterization of the central cooling channel for the DEMO PF coil conductor

IF 2 3区工程技术 Q1 NUCLEAR SCIENCE & TECHNOLOGY

Fusion Engineering and Design Pub Date : 2025-06-20 DOI:10.1016/j.fusengdes.2025.115258

Aleksandra Dembkowska , Monika Lewandowska , Paweł Herbin

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Abstract

Recently a design of the 6 Poloidal Field (PF) coils for the European DEMOnstration Fusion Power Plant (EU-DEMO) tokamak has been proposed by the CEA IRFM team. Each of PF coils designed by CEA is wound using a square NbTi Cable-in-Conduit Conductor (CICC) with a central cooling channel separated by a steel spiral from the bundle region. The outer diameters of spirals in the EU-DEMO PF3, PF4 and PF6 conductors designed by CEA are in the range 13.6 - 15.8 mm. Spirals with such large diameters have never been tested for pressure drop, so applicability of the available friction factor correlations for them is uncertain. To clarify this issue, we produced (by 3D printing) two samples relevant for hydraulic tests, with geometry similar to spiral cooling channels in CICCs. The first one, with the geometry identical to the ITER SHOWA spiral which had been tested for pressure drop in the past, served as verification, if samples produced by 3D printing feature the same hydraulic resistance as the respective samples traditionally produced (by inserting a spiral into a tight steel or rubber tube). The second sample had a large diameter identical to the cooling channel of the PF6 conductor designed by CEA. Both samples were tested for pressure drop at the THETIS installation, using demineralized water at three different temperatures, to get experimental data in possibly wide Re range. The experimental data were used to develop friction factor correlation for a spiral duct with large diameter, which could be utilized in future thermal-hydraulic analyses of the EU-DEMO PF coils or other CICCs with cooling channels of similar geometry.

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DEMO PF线圈导体中央冷却通道的水力特性

最近，CEA的IRFM团队提出了一种用于欧洲示范核聚变发电厂（EU-DEMO）托卡马克的6个极向场（PF）线圈的设计方案。CEA设计的每个PF线圈都使用方形的NbTi电缆导管导体（CICC）绕线，中心冷却通道由钢螺旋与束区分开。CEA设计的EU-DEMO PF3、PF4和PF6导体的螺旋外径在13.6 ~ 15.8 mm之间。如此大直径的螺旋从未测试过压降，因此可用的摩擦系数相关性对它们的适用性是不确定的。为了澄清这个问题，我们（通过3D打印）制作了两个与水力测试相关的样品，其几何形状类似于ccic中的螺旋冷却通道。第一个，与过去测试过压降的ITER SHOWA螺旋的几何形状相同，作为验证，如果3D打印生产的样品与传统生产的样品具有相同的水力阻力（通过将螺旋插入紧钢或橡胶管）。第二种样品直径较大，与CEA设计的PF6导体冷却通道相同。两种样品都在THETIS装置上进行了压降测试，使用三种不同温度的脱盐水，以获得可能较宽的Re范围内的实验数据。利用实验数据建立了大直径螺旋风道的摩擦系数相关性，可用于EU-DEMO PF线圈或其他具有类似几何形状冷却通道的ccic的热水力分析。

本文章由计算机程序翻译，如有差异，请以英文原文为准。

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来源期刊

Fusion Engineering and Design 工程技术-核科学技术

CiteScore

3.50

自引率

23.50%

发文量

275

审稿时长

3.8 months

期刊介绍： The journal accepts papers about experiments (both plasma and technology), theory, models, methods, and designs in areas relating to technology, engineering, and applied science aspects of magnetic and inertial fusion energy. Specific areas of interest include: MFE and IFE design studies for experiments and reactors; fusion nuclear technologies and materials, including blankets and shields; analysis of reactor plasmas; plasma heating, fuelling, and vacuum systems; drivers, targets, and special technologies for IFE, controls and diagnostics; fuel cycle analysis and tritium reprocessing and handling; operations and remote maintenance of reactors; safety, decommissioning, and waste management; economic and environmental analysis of components and systems.