HHF Tests Research Articles

Beryllium dust generation caused by HHF tests

Beryllium will be used as a plasma-facing material of the ITER reactor components. One such component is the ITER first-wall panel, which contains a layer of beryllium tiles on the plasma-facing surface. Full-Scale Prototypes (FSP) of these panels are undergoing qualification through a series of cyclic High-Heat-Flux (HHF) tests at the HELCZA (High Energy Load Czech Assembly) facility. Such process, directly and indirectly, causes disintegration of the beryllium surface and results in beryllium dust formation. With regards to the beryllium toxicity, it is crucial to monitor the occurrence and the spread of the mentioned dust to ensure personnel safety. Moreover, it is necessary to decontaminate FSPs from the beryllium dust at the end of HHF tests. This study deals with the monitoring of the beryllium dust concentration levels on the FSP after HHF testing at HELCZA and its spread to the vicinity of the testing area, with the aim to provide a basis for safe operation.

Thermomechanical Analysis of a PFC Integrating W Lattice Armour in Response to Different Plasma Scenarios Predicted in the EU-DEMO Tokamak

Despite the high performance exhibited by tungsten (W), no material would be able to withstand the huge loads expected with extreme plasma transients in EU-DEMO and future reactors, where the installation of sacrificial first wall limiters is essential to prevent excessive wall degradation. The integration of W lattices in the architecture of such components can allow for meeting their conflictual requirements: indeed, they must ensure the effective exhaust of the nominal thermal load during stationary operation; when transients occur, they must thermally insulate and decouple the surface from the heat sink, promoting prompt vapour shielding formation. Starting from the optimised layouts highlighted in a previous study, in this work, a detailed 3D finite element model was developed to analyse in depth the influence of the actual features of the latticed metamaterial on the overall performance of the EU-DEMO limiter PFC on the basis of a flat tile configuration. Its main goal is to help in identifying the most promising layout as a preconceptual design for the fabrication of a small-scale mock-up. For this purpose, the complex geometry of a W-based lattice armour was faithfully reproduced in the model and analysed. This allowed for a detailed assessment of the thermally induced stresses that develop in the component because of the temperature field in response to a number of plasma scenarios—above all, normal operation and ramp down. Structural integrity was verified through the acceptance criteria established for ITER. The two optimised layouts proposed for the PFC were able to effectively meet the requirements under normal reactor operating conditions, while they missed some requirements in the ramp-down case. However, the first HHF tests will be performed in order to benchmark the analyses.

A method for high heat flux fatigue test of W/Cu/steel small mock-up using normal cooling conditions

• Amethodfor HHF test of W/steel small mockup is introduced, using normal coolant conditions. • Fatigue life of W/Cu/steel mockup is preliminary predicted and tested. • The experiment result shows consistency with the predicted life. Fatigue property of W/steel joints is important for predicting the life of First Wall (FW). Few facilities can satisfy the harsh coolant parameters of FW. It is necessary to develop a method to evaluate the fatigue property, using coolant with normal pressure and temperature, for example 0.2 MPa and 25 ℃. Combining numerical calculation and experiment validation. Building the relationship between different heat flux and the maximum temperature and strain, predicting the life of FW according to Manson-Coffin formula (MC). Verifying by the fatigue test in a High Heat Flux test platform equipped with an electron gun, the experimental results have shown the feasibility of this method.

Analysis of hardness and microstructural changes in Tungsten mono-blocks exposed to high heat flux at 10 MW/m2

In this study, we investigated the material properties of tungsten mono-blocks undergone high heat flux test. The ITER-like tungsten mono-block specimens were fabricated by two kinds of joining methods. The tungsten-Cu interlayer joint was made by gas pressure casting and Cu interlayer-CuCrZr cooling tube were joined by hot radial pressing (HRP) or brazing method. To evaluate the integrity of the fabricated mono-blocks, high heat flux tests were done on the small mockups under cyclic heat load of 10 MW/m2 up to 5,000 cycles. We measured the microstructure, hardness, and surface state of the tungsten armors before and after the heat flux exposure experiment. The Vickers hardness profile measured from the surface down to the Cu interlayer showed a noticeable reduction around 200 μm ~ 500 μm depth after high heat flux test compared to that before the test depending on the bonding quality of mockup specimens. At depth beyond 800 μm, the Vickers hardness did not show any notable difference before and after the high heat flux test, which indicates the material property of tungsten is not so much affected in this region at the above heat flux condition. The microstructure and grain size distribution was measured and analyzed by SEM and EBSD. The correlation between the microstructure and hardness profile, and the impacts of the mockup performance under HHF test to the integrity of tungsten armors are analyzed.

High-heat-flux technologies for the European demo divertor targets: State-of-the-art and a review of the latest testing campaign

Divertor target is one of the most critical in-vessel components in a fusion power plant being in charge of particle and power exhaust. The targets are exposed to severe thermal loads produced by steady bombardment of impinging plasma flux. Since 2014, integrated R&D efforts have been continued aiming at developing a design concept and high-heat-flux (HHF) technologies for divertor targets of the European DEMO reactor. Recently, the second round (2017–2019) of the R&D program has been concluded. As in the first R&D round, five water-cooled target design concepts were further developed and evaluated. Fabrication technologies were improved reaching a consolidated production quality. Extensive HHF tests were conducted using small-scale mock-ups for extended loading regimes (heat flux: 20–32MW/m²). Comparative studies were performed to investigate effects of copper interlayer thickness (0.1–1 mm) and different tungsten armor materials. In the present paper, the final results of the second round HHF testing campaign are reported. The HHF performance of each design variant is discussed based on in-situ diagnostic data (infrared thermography), ultrasonic inspection images and postmortem metallographic micrographs. All monoblock-type design concepts passed the specified qualification criterion (≥500 pulses at 20 MW/m², coolant: 130 °C) without any failure or armor cracking. Moreover, two of them (ITER-like and composite pipe) remained fully intact even under 25 MW/m² (100 pulses) and 32 MW/m² (5 pulses).

Performance assessment of thick W/Cu graded interlayer for DEMO divertor target

Abstract The development of a divertor target for DEMO is of great importance, being able to sustain the harsh environment that is imposed on this component. To fulfill the loading requirements, different concepts were developed within the EUROfusion WPDIV project. The baseline concept is based on the ITER divertor target W-monoblock design. It is made of tungsten as armour material, CuCrZr as structural material and Cu-OFHC as compliant layer. One of the proposed alternative concepts aims to minimize the stress at interfaces by replacing the thick copper interlayer by W/Cu functionally graded material (FGM). In this study, the FGM interlayer, with a thickness of 500 μm, is composed of stacked elementary layers with the following three compositions: 25 vol.%W + 75 vol.%Cu, 50 vol.%W + 50 vol.%Cu, 75 vol.%W + 25 vol.%Cu. Several FGM interlayers were studied. In total three monoblock type mock-ups were manufactured. This paper describes the steps needed to manufacture mock-ups and characterization of elementary layers (composition, porosity, Young’s modulus). HHF tests applying up to 1000 cycles at 20 MW/m² and sub-sequent post-mortem examinations were performed to qualify the concept performance.

Assessment of the high heat flux performance of European DEMO divertor mock-ups

In the framework of the DEMO divertor project an extensive R&D programme has been carried out to develop advanced design concepts for hot water cooled divertor targets. These plasma-facing components are made of W monoblocks as plasma facing material bonded on CuCrZr cooling tubes. They are designed to allow a reliable DEMO operation of 2 h long pulses up to a 20 MW m−2 maximum heat flux. Considering the current experiences of the qualification of ITER divertor components, the operation at significantly higher neutron fluence, the longer required lifetime, and the coolant temperature of 150 °C are challenges for design and manufacturing. The aim of the performed HHF tests is to support the identification of the most promising design concepts for further development. Eight different types of monoblock mock-ups were designed and manufactured by the involved research groups. These mock-ups were assessed by HHF examination in the test facility GLADIS at IPP Garching. The HHF tests were focused on 20 MW m−2 loading with 500–1000 cycles using hot water cooling to simulate slow thermal transients in DEMO. In screening tests heat fluxes up to 32 MW m−2 were applied to study the performance of the components against single thermal events exceeding the expected heat flux limits. This paper gives an overview of the main results obtained from the HHF tests of about 45 components. The microscopic examination after loading analyses recrystallization effects and investigates the W/Cu interfaces in more detail.

Progress in high heat flux testing of European DEMO divertor mock-ups

In the framework of the DEMO divertor project of EUROfusion an extensive R&D program has been carried out to develop advanced design concepts for hot water cooled divertor targets. These plasma-facing components made of W blocks as plasma facing material and CuCrZr as cooling tubes should allow a reliable DEMO operation for 2 h long pulses and maximum heat fluxes up to 20 MW/m². Compared to ITER, the operation at the higher coolant temperature of 150 °C, the longer required lifetime, and the significantly higher neutron fluence are the design challenges exceeding the current extent of experience. In this study we present high heat flux test results of a total of 34 tested W monoblock mock-ups. 14 of them where tested with 20 MW/m² heat flux at DEMO hot water cooling conditions up to now. All tests were performed in the HHF test facility GLADIS from 2016 to 2018. Furthermore, we discuss results from the post exposure investigation of selected concepts of the first R&D phase.

ITER-like divertor target for DEMO: Design study and fabrication test

As a major in-vessel component of a tokamak-type fusion reactor, the divertor is mainly in charge of removal of particles and partial power exhaust via scrape-off layer. The target plate of the divertor is directly exposed to non-uniform high heat flux on the surface by particle bombardment and radiation. In the case of ITER and a DEMO reactor, the peak surface heat flux is expected to reach up to 10MW/m2 during normal operation and 20MW/m2 during slow transient events like loss of plasma detachment.This paper reports the results of a preliminary code-based design study and fabrication technology verification test which were conducted for developing an ITER-like divertor target design for the DEMO divertor.The structural failure evaluation against the ratchetting and fatigue criteria of the ITER SDC-IC showed that the design with reduced dimensions would allow sufficient design margin (reserve factor) for three distinct thermal loading cases. The first trial of mock-up fabrication using a new joining furnace at ENEA was successfully completed. The ultrasonic inspection test made before and after the cyclic HHF tests at GLADIS facility demonstrated high quality of fabrication and robust design concept.

Results of high heat flux qualification tests of W monoblock components for WEST

One goal of the WEST project (W Environment in Steady-state Tokamak) is the manufacturing, quality assessment and operation of ITER-like actively water-cooled divertor plasma facing components made of tungsten. Six W monoblock plasma facing units (PFUs) from different suppliers have been successfully evaluated in the high heat flux test facility GLADIS at IPP. Each PFU is equipped with 35 W monoblocks of an ITER-like geometry. However, the W blocks are made of different tungsten grades and the suppliers applied different bonding techniques between tungsten and the inserted Cu-alloy cooling tubes. The intention of the HHF test campaign was to assess the manufacturing quality of the PFUs on the basis of a statistical analysis of the surface temperature evolution of the individual W monoblocks during thermal loading with 100 cycles at 10 MW m−2. These tests confirm the non-destructive examinations performed by the manufacturer and CEA prior to the installation of the WEST platform, and no defects of the components were detected.

Manufacturing and testing of self-passivating tungsten alloys of different composition

Self-passivating tungsten based alloys for the first wall armour of future fusion reactors are expected to provide a major safety advantage compared to pure tungsten in case of a loss of coolant accident with simultaneous air ingress, due to the formation of a stable protective scale at high temperatures in presence of oxygen which prevents the formation of volatile and radioactive WO3.Bulk W-15Cr, W-10Cr-2Ti and W-12Cr-0.5Y alloys were manufactured by mechanical alloying followed by can encapsulation and HIP. This route resulted in fully dense materials with nano-structured grains. The ability of Ti and especially of Y to inhibit grain growth was observed in the W-10Cr-2Ti and W-12Cr-0.5Y alloys. Besides, Y formed Y-rich oxide nano-precipitates at the grain boundaries, and is thus expected to improve the mechanical behaviour of the Y-containing alloy. Isothermal oxidation tests at 800 ºC (1073K) and oxidation tests under accident-like conditions revealed that the W-12Cr-0.5Y alloy exhibits the best oxidation behaviour of all alloys, especially in the accident-like scenario. Preliminary HHF tests performed at GLADIS indicated that the W-10Cr-2Ti alloy is able to withstand power densities of 2 MW/m2 without significant damage of the bulk structure. Thermo-shock tests at JUDITH-1 to simulate mitigated disruptions resulted in chipping of part of the surface of the as-HIPed W-10Cr-2Ti alloy. An additional thermal treatment at 1600 °C (1873K) improves the thermo-shock resistance of the W-10Cr-2Ti alloy since only crack formation is observed.

High heat flux test of tungsten brazed mock-ups developed for KSTAR divertor

The tungsten (W) brazed flat type mock-up which consists of W, OFHC-Cu (oxygen-free high conductive copper) and CuCrZr alloy has been designed for KSTAR divertor in preparation for KSTAR upgrade with 17MW heating power. For verification of the W brazed mock-up, the high heat flux test is performed at KoHLT-EB (Korea High Heat Load Test Facility-Electron Beam) in KAERI (Korea Atomic Energy Research Institute). Three mock-ups are tested for several thousand thermal cycles with absorbed heat flux up to 5MW/m2 for 20s duration. There is no evidence of the failure at the bonding joints of all mock-ups after HHF test. Finite element analysis (FEA) is performed to interpret the result of the test. As a result, it is considered that the local area in the water is in the subcooled boiling regime.

Progress on performance assessment of ITER enhanced heat flux first wall technology after neutron irradiation

ITER first wall (FW) panels are irradiated by energetic neutrons during the nuclear phase. Thus, an irradiation and high heat flux testing programme is undertaken by the ITER organization in order to evaluate the effects of neutron irradiation on the performance of enhanced heat flux (EHF) FW components. The test campaign includes neutron irradiation (up to 0.6–0.8 dpa at 200 °C–250 °C) of mock-ups that are representative of the final EHF FW panel design, followed by thermal fatigue tests (up to 4.7 MW m−2). Mock-ups were manufactured by the same manufacturing process as proposed for the series production. After a pre-irradiation thermal screening, eight mock-ups will be selected for the irradiation campaigns. This paper reports the preparatory work of HHF tests and neutron irradiation, assessment results as well as a brief description of mock-up manufacturing and inspection routes.

High heat flux testing of divertor plasma facing materials and components using the HHF test facility at IPR

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.

He-Cooled Divertor for DEMO: Technological Study on Joining Tungsten Components with Titanium Interlayer

A modular He-cooled divertor concept for DEMO has been pursued at KIT with the goal of reaching 10 MW/m2. The reference design uses small tungsten-based cooling fingers of about 20 mm in size. They consist of a tungsten tile as a thermal shield that is to be connected to a thimble heat sink structure from W–1 wt% La2O3 (WL10) tungsten alloy. The lower boundary of the divertor operating temperature window is predicted by the ductile-to-brittle temperature and the upper boundary by the recrystallization temperature of WL10 material, currently assumed at 600°C and 1300°C, respectively. The important requirements for the joint between the W tile and WL10 thimble are (a) functioning as a crack stopper, (b) resisting a high operating temperature of about 1200°C, and (c) using low-activation material as an interlayer. Previously used PdNi brazing material has been successfully tested at a brazing temperature of about 1270°C. The mock-ups produced in this way are sufficient for the HHF tests without neutrons. In a further step to approach the DEMO requirements with higher demands, the use of low-activating titanium with a melting point of 1668°C as bonding material was examined both for brazing and for diffusion welding of tungsten parts. This paper reports on the first successful test results of both high-temperature brazing and diffusion bonding techniques.

Fabrication and Integrity Testing of a Korean ITER TBM FW Mock-Up in Preparation for High Heat Flux Test

Korea (KO) has developed and participated in the Test Blanket Module (TBM) program in the ITER, in which Ferritic Martensitic (FM) steel was used as the structural material for the TBM first wall (FW). To develop the fabrication method for the TBM FW and verify its integrity, a half-scale sub-module mock-up was fabricated and integrity test has been prepared; A dimension of it is a 444 mm height, 260 mm width, and 435 mm depth. A mock-up was assembled by HIPping of the previous fabricated components between the welded front and back plates, and then machining these plates to form the completed sub-module FW. To investigate the deformation of the cooling channels after the fabrication of the mock-up, neutron radiography was carried using neutron radiography facility. Pressure and He leak tests were successfully performed without any leak and failure. The flow rates in each channel were measured with the conventional ultrasonic sensor but it shows non-uniform flow distribution at each channel differently from the estimation by ANSYS-CFX. HHF test conditions were evaluated through the ANSYS-CFX analysis considering the above measured flow rates in each channel and it shows non-uniform temperature distribution of the FW mock-up. We will re-assemble or fabricate the manifold and perform the flow test before HHF test in the near future.

Conceptual Design of Vacuum Chamber for testing of high heat flux components using electron beam as a source

A conceptual design of vacuum chamber is proposed to study the thermal response of high heat flux components under energy depositions of the magnitude and durations expected in plasma fusion devices. It is equipped with high power electron beam with maximum beam power of 200 KW mounted in a stationary horizontal position from back side of the chamber. The electron beam is used as a heat source to evaluate the heat removal capacity, material performance under thermal loads & stresses, thermal fatigue etc on actively cooled mock – ups which are mounted on a flange system which is the front side door of the chamber. The tests mock – ups are connected to a high pressure high temperature water circulation system (HPHT-WCS) operated over a wide range of conditions. The vacuum chamber consists of different ports at different angles to view the mock -up surface available for mock -up diagnostics. The vacuum chamber is pumped with different pumps mounted on side ports of the chamber. The chamber is shielded from X – rays which are generated inside the chamber when high-energy electrons are incident on the mock-up. The design includes development of a conceptual design with theoretical calculations and CAD modelling of the system using CATIA V5. These CAD models give an outline on the complete geometry of HHF test chamber, fabrication challenges and safety issues. FEA analysis of the system has been performed to check the structural integrity when the system is subjected to structural & thermal loads.

Progress of He-cooled divertor development for DEMO

A He-cooled divertor concept for DEMO [1] has been developed at Karlsruhe Institute of Technology (KIT) since a couple of years with the goal of reaching a heat flux of 10 MW/m 2 anticipated for DEMO. The reference concept HEMJ (He-cooled modular divertor with multiple-jet cooling) is based on the use of small cooling fingers – each composed of a tungsten tile brazed to a tungsten alloy thimble – as well as on impingement jet cooling with helium at 10 MPa, 600 °C. The cooling fingers are connected to the main structure of ODS Eurofer steel by brazing in combination with a mechanical interlock. This paper reports progress to date of the design accompanying R&Ds, i.e. primarily the fabrication technology and HHF experiments. For the latter a combined helium loop and electron beam facility (200 kW, 40 keV) at Efremov Institute, St. Petersburg, Russia, has been used. This facility enables mock-up testing at a nominal helium inlet temperature of 600 °C, a pressure of 10 MPa, and a maximal pressure head of 0.5 MPa. HHF test results till now confirm well the divertor design performance. In the recent test series in early 2010 the first breakthrough was achieved when a mock-up has survived over 1000 cycles at 10 MW/m 2 unscathed.

He-cooled divertor for DEMO: Status of development and HHF tests

A modular He-cooled divertor concept for DEMO has been pursued at Karlsruhe Institute of Technology (KIT) in cooperation with Efremov with the goal of reaching a DEMO-relevant heat flux of 10 MW/m 2 and 1000 load cycles. The reference design is based on the use of small cooling finger modules made of tungsten and tungsten alloy and on impingement cooling with helium. For the purpose of experimental verification of the design, a combined helium loop and electron beam facility was built at Efremov. Technological studies on fabrication of tungsten divertor mock-ups were carried out. Results of the first HHF test series 2006 confirmed already the divertor performance of 10 MW/m 2. After some improvements of mock-ups the results of the following second and third test series in 2007–2008 showed noticeable enhancement in performance and resistance against thermal cyclic loadings. The fourth test series was continued in early 2010 after an upgrade of the EB facility. Six surviving mock-ups from the last test series were further tested. They survived differently between 180 and 1100 cycles in total under the maximum heat load of at least 10 MW/m 2 before failure. Thus the first breakthrough is reached.

Qualification of high heat flux components: application to target elements of W7-X divertor

The development of actively cooled plasma-facing components (PFC) represents one of fusion's most challenging engineering efforts. In this frame, a high-quality bonding between the refractory armour and the heat sink is essential to ensure the heat removal capability and the thermal performances of PFC. Experience gained during manufacturing of Tore Supra actively cooled PFC led to the establishment of a qualification methodology and provided a large experience of acceptance criteria using an active infrared thermography (système d'acquisition de traitement infra-rouge, SATIR). This paper presents the application of this qualification process to the W7-X pre-series components, with the objective of assessing and defining workable acceptance criteria that enable reliable predictions of performance at the nominal heat flux requirements in W7-X. Finally, to check the reliability of the non-destructive examination (NDE) method by transient infrared thermography, the newly defined acceptance criteria were applied to W7-X pre-series target elements (batch #3). The SATIR results, benchmarked with HHF tests performed on the GLADIS ion beam facility were discussed to assess the ability to detect critical defects at the interface between tiles and heat sink.