Manufacturing and high heat flux testing of advanced target mock-ups for the EU-DEMO divertor target

This paper is focused on various aspects of the development and testing of water cooled divertor PFCs. Divertor PFCs are mainly designed to absorb the heat and particle fluxes outflowing from the core plasma of fusion devices like ITER. The Divertor and First Wall Technology Development Division at the Institute for Plasma Research (IPR), India, is extensively working on development and testing of divertor plasma facing components (PFCs). Tungsten and graphite macro-brush type test mock-ups were produced using vacuum brazing furnace technique and tungsten monoblock type of test mock-ups were obtained by hot radial pressing (HRP) technique. Heat transfer performance of the developed test mock-ups was tested using high heat flux tests with different heat load conditions as well as the surface temperature monitoring using transient infrared thermography technique. Recently we have established the High Heat Flux Test Facility (HHFTF) at IPR with an electron gun EH300V (M/s Von Ardenne Anlagentechnik GmbH, Germany) having maximum power 200 kW. Two tungsten monoblock type test mock-ups were probed using HHFTF. Both of the test mock-ups successfully sustained 316 thermal cycles during high heat flux (HHF) tests. The test mock-ups were non-destructively tested using infrared thermography before and after the HHF tests. In this note we describe the detailed procedure used for testing macro-brush and monoblock type test mock-ups using in-house transient infrared thermography set-up. An acceptance criteria limit was defined for small scale macro-brush type of mock-ups using DTrefmax value and the surface temperature measured during the HHF tests. It is concluded that the heat transfer behavior of a plasma facing component was checked by the HHF tests followed by transient IR thermography. The acceptance criteria DTrefmax limit for a graphite macro-brush mock-up was found to be ~3°C while for a tungsten macro-brush mock-up it was ~5°C.

The Institute for Plasma Research (IPR) is involved in the development of various technologies for fabrication and testing/qualification of tungsten (W)–based plasma-facing components for ITER-like applications. As a part of these activities, a small-scale W/Cu/CuCrZr monoblock test mock-up is prepared by using a vacuum brazing process. Further, thermal fatigue testing of this test mock-up is performed using the High Heat Flux Test Facility at IPR under an absorbed heat flux of ~9.5 MW/m2 for 1200 thermal cycles. Thermal-hydraulic analysis and computational fluid dynamics (CFD) simulations are also performed to understand the heat flux testing scenario. Nondestructive testing of the test mock-up after thermal cyclic heat flux testing indicates that there are no defects in the test mock-up except for minor cracks on the surface of W tiles. The results of thermal cyclic testing, finite element analysis, CFD analysis, and postanalysis of the High Heat Flux test are presented in the paper.

Manufacturing and testing of flat-type divertor mockup with advanced materials

The High Heat Flux Test Facility (HHFTF) was designed and established recently at Institute for Plasma Research (IPR) in India for testing heat removal capability and operational life time of plasma facing materials and components of the ITER-like tokamak. The HHFTF is equipped with various diagnostics such as IR cameras and IR-pyrometers for surface temperature measurements, coolant water calorimetry for absorbed power measurements and thermocouples for bulk temperature measurements. The HHFTF is capable of simulating steady state heat load of several MW m−2 as well as short transient heat loads of MJ m−2. This paper presents the current status of the HHFTF at IPR and high heat flux tests performed on the curved tungsten monoblock type of test mock-ups as well as transient heat flux tests carried out on pure tungsten materials using the HHFTF. Curved tungsten monoblock type of test mock-ups were fabricated using hot radial pressing (HRP) technique. Two curved tungsten monoblock type test mock-ups successfully sustained absorbed heat flux up to 14 MW m−2 with thermal cycles of 30 s ON and 30 s OFF duration. Transient high heat flux tests or thermal shock tests were carried out on pure tungsten hot-rolled plate material (Make:PLANSEE) with incident power density of 0.49 GW m−2 for 20 milliseconds ON and 1000 milliseconds OFF time. A total of 6000 thermal shock cycles were completed on pure tungsten material. Experimental results were compared with mathematical simulations carried out using COMSOL Multiphysics for transient high heat flux tests.

Overview of the Japanese mock-up tests for ITER high heat flux components

Plasma-facing components (PFCs) usually need to withstand extreme incident heat flux conditions in nuclear fusion engineering. In situ measurements of deformation to fatigue failure of PFCs under high heat flux (HHF) tests are significant and essential to understand their thermal-mechanical behaviors under servicing conditions; these measurements can provide first-hand and important information for evaluating and optimizing design performance and manufacturing techniques. However, unlike traditional contact measurements with strain gauges, the contactless optical measurement technique is rarely applied to measure or monitor the deformation to fatigue damage process of PFCs employed in HHF tests. In this work, a comprehensive HHF experimental platform was established by combining an electron gun scanning HHF heating system with a three-dimensional digital image correlation (3D-DIC) measurement system based on a vacuum chamber and specially designed optical windows and light sources. The 3D-DIC technology was utilized to measure the field deformation of a divertor mockup under cyclic HHF loads. Validation and qualification tests were conducted on the comprehensive experimental platform to ensure the performance of the platform and the accuracy of the 3D-DIC method. The thermal-induced field deformation of a flat-type divertor mockup under the conditions of 1000 HHF cycles at 10 MW m−2 and 300 HHF cycles at 20 MW m−2 using the 3D-DIC technique was then measured. The mechanical behavior of the accumulated plastic strain and fatigue debonding failure of the W/Cu interface due to periodic thermal stress were captured in situ by the measured strain curves and contours for the first time during HHF tests. The results demonstrate the feasibility and accuracy of the 3D-DIC technique for in situ fatigue deformation and damage strain measurements of PFCs during HHF tests. The proposed methods and technologies are expected to be applied to measure and monitor the servicing performance of PFCs under servicing conditions.

Design and development of LN2 cooled cryopump for application in high heat flux test facility

Korea Domestic Agency (KO-DA) was responsible for the procurement of the ITER blanket modules 1, 2, and 6 in the original procurement allocation. According to the procurement reallocation of the blanket system, Korea will procure the blanket shield block in place of the blanket first wall. Nevertheless, several R&D activities in Korea have been performed including optimization of the hot isostatic pressing (HIP) bonding process between Be/CuCrZr and CuCrZr/SS, the nondestructive test method, fabrication feasibility study, high heat flux tests, and design analysis. Especially, KO-DA participated in the qualification program for the mock-up manufacture and high heat flux tests. Several mock-ups were fabricated and tested during the qualification program. The details of the mock-up manufacture and test results are described in this paper. Also, two heat flux facilities were installed based on the graphite heating, and a new electron beam heat flux facility will be built in the near future for the enhanced heat flux mock-up test. As well, some design analysis was performed to investigate the performance of the blanket first wall against thermo-mechanical loading. In this paper, the status of the R&D activities and the results of the qualification tests for KO mock-ups are reviewed.

Performance of straight tungsten mono-block test mock-ups using new high heat flux test facility at IPR

The divertor target components for the Chinese fusion engineering test reactor (CFETR) and the future experimental advanced superconducting tokamak (EAST) need to remove a heat flux of up to ∼20 MW m−2. In view of such a high heat flux removal requirement, this study proposes a conceptual design for a flat-tile divertor target based on explosive welding and brazing technology. Rectangular water-cooled channels with a special thermal transfer structure (TTS) are designed in the heat sink to improve the flat-tile divertor target’s heat transfer performance (HTP). The parametric design and optimization methods are applied to study the influence of the TTS variation parameters, including height (H), width (W*), thickness (T), and spacing (L), on the HTP. The research results show that the flat-tile divertor target’s HTP is sensitive to the TTS parameter changes, and the sensitivity is T > L > W* > H. The HTP first increases and then decreases with the increase of T, L, and W* and gradually increases with the increase of H. The optimal design parameters are as follows: H = 5.5 mm, W* = 25.8 mm, T = 2.2 mm, and L = 9.7 mm. The HTP of the optimized flat-tile divertor target at different flow speeds and tungsten tile thicknesses is studied using the numerical simulation method. A flat-tile divertor mock-up is developed according to the optimized parameters. In addition, high heat flux (HHF) tests are performed on an electron beam facility to further investigate the mock-up HTP. The numerical simulation calculation results show that the optimized flat-tile divertor target has great potential for handling the steady-state heat load of 20 MW m−2 under the tungsten tile thickness <5 mm and the flow speed ≥7 m s−1. The heat transfer efficiency of the flat-tile divertor target with rectangular cooling channels improves by ∼13% and ∼30% compared to that of the flat-tile divertor target with circular cooling channels and the ITER-like monoblock, respectively. The HHF tests indicate that the flat-tile divertor mock-up can successfully withstand 1000 cycles of 20 MW m−2 of heat load without visible deformation, damage, and HTP degradation. The surface temperature of the flat-tile divertor mock-up at the 1000th cycle is only ∼930 °C. The flat-tile divertor target’s HTP is greatly improved by the parametric design and optimization method, and is better than the ITER-like monoblock and the flat-tile mock-up for the WEST divertor. This conceptual design is currently being applied to the engineering design of the CFETR and EAST flat-tile divertors.

High heat flux test results for a thermal break DEMO divertor target and subsequent design and manufacture development

Fabrication and high heat flux test with the first wall mockups for developing the KO TBM

The back plate is an important component of the ion source because of its multiple roles including heat load removal during beam operation. The main components of the back plate are (1) a Type 304L stainless steel (SS304L) magnet positioning plate that holds samarium cobalt permanent magnets required for the confinement of ion source plasma, (2) an oxygen-free electronic copper cooling plate with 35 inner and 8 outer cooling channel grooves (each of which is 4 × 1.8 mm2) that is vacuum brazed with a SS304L magnet positioning plate, and (3) a SS304L magnet cover plate. In this paper, the back plate is successfully fabricated, and a high heat flux experiment is done at the High Heat Flux Test Facility Center with an electron beam power of 200 kW for 458 s. The uniform incident heat flux is 2.5 MW/m2. Demineralized water at 34°C is supplied at the rate of 1 kg/s to the cooling plate at inlet pressure of 8.2 bars to remove the high heat load. The surface temperature of the copper plate is measured by an infrared camera, and three temperature regions are observed. The measured average surface temperature of the cooling plate is ~152°C. The bulk water temperature rise ΔTw is ~39.42°C. The estimated absorbed heat flux is ~2.04 MW/m2, and the heat absorption coefficient is 81.6%. The measured leak rate after the heat flux test is 1.6 × 10−8 mbars∙L/s. These High Heat Flux Test experimental results will be useful to study the thermomechanical performance of the back plate and to understand the effect of increasing the beam pulse length.

He-cooled divertor development for DEMO

Non-destructive testing of CFC monoblock divertor mock-ups