Dennis L. Youchison

Critical heat flux analysis and R&D for the design of the ITER divertor

The vertical target and dump target of the ITER divertor have to be designed for high heat fluxes (up to 20 MW:m 2 over :10 s). Accommodation of such high heat fluxes gives rise to several issues, including the critical heat flux (CHF) margin which is a key requirement influencing the choice of cooling channel geometry and coolant conditions. An R&D programme was evolved to address the overall CHF issue and to help focus the design. It involved participation of the four ITER home teams and has been very successful in substantially expanding the CHF data base for one-sided heating and in providing more accurate experimental measurements of pressure drop (and derived correlations) for these geometries. This paper describes the major R&D results and the design analysis performed in converging on a choice of reference configuration and parameters which resulted in a CHF margin of : 1.4 or more for all divertor components.

Tritium Inventory in the ITER PFC's: Predictions, Uncertainties, R&D Status and Priority Needs

Abstract New data on hydrogen plasma isotopes retention in beryllium and tungsten are now becoming available from various laboratories for conditions similar to those expected in the International Thermonuclear Experimental Reactor (ITER) where previous data were either missing or largely scattered. Together with a significant advancement in understanding, they have warranted a revisitation of the previous estimates of tritium inventory in ITER, with beryllium as the plasma facing material for the first-wall components, and tungsten in the divertor with some carbon-fibre-composites clad areas, near the strike points. Based on these analyses, it is shown that the area of primary concern, with respect to tritium inventory, remains codeposition with carbon and possibly beryllium on the divertor surfaces. Here, modelling of ITER divertor conditions continues to show potentially large codeposition rates which are confirmed by tokamak findings. Contrary to the tritium residing deep in the bulk of materials, this surface tritium represents a safety hazard as it can be easily mobilised in the event of an accident. It could, however, be possibly removed and recovered. It is concluded that active and efficient methods to remove the codeposited layers are needed in ITER and periodic conditioning/cleaning would be required to control the tritium inventory and avoid exhausting the available fuel supply. Some methods which could possibly be used for in-situ cleaning are briefly discussed in conjunction with the research and development work required to extrapolate their applicability to ITER.

Performance of the different tungsten grades under fusion relevant power loads

Abstract The test results of several W grades at conditions typical for plasma facing component operations are summarised. These include the effects of steady-state heat fluxes (up to 43 MW / m 2 ), disruption simulation (up to 30 MJ / m 2 during 0.05–0.36 ms) and heat flux tests of W after disruption simulation. Representatives of the main W grades have been investigated: pure sintered W, W–Re and W–Mo cast alloys, W –1% La 2 O 3 , W –2% CeO 2 , single crystal W, etc. The resistance to high heat fluxes strongly depends on the orientation of the W grains to incident heat flux and with proper orientation W can withstand heat fluxes up to 27 MW / m 2 . After disruption simulation, intensive surface crack formation has been observed for all studied W grades except single crystal W. Severe damage after disruption and thermal fatigue loading have been observed for almost all W grades except the W–5Re–0.1ZrC alloy and W–Re single crystal.

The structure, properties and performance of plasma-sprayed beryllium for fusion applications

Plasma-spray technology is under investigation as a method for producing high thermal conductivity beryllium coatings for use in magnetic fusion applications. Recent investigations have focused on optimizing the plasmaspray process for depositing beryllium coatings on damaged beryllium surfaces. Of particular interest has been optimizing the processing parameters to maximize the through-thickness thermal conductivity of the beryllium coatings. Experimental results will be reported on the use of secondary H2 gas additions to improve the melting of the beryllium powder and negative transferred-arc cleaning to improve the bonding between the beryllium coatings and the underlying surface. Information will also be presented on thermal cycle tests which were done on beryllium coated ISX-B beryllium limiter tiles using 10s cycle times with 60s cooldowns using a heat flux slightly in excess of 5 MW/m2.

Thermal performance and flow instabilities in a multi-channel, helium-cooled, porous metal divertor module

Abstract Pressurized helium is under consideration for cooling Langmuir probes and plasma facing components of next generation fusion experiments. Helium is non-corrosive, does not activate, separated easily from tritium, vacuum compatible, and undergoes no phase transformations. Recently, the thermal performance of a bare-copper, dual-channel, helium-cooled, porous metal divertor mock-up, designed and fabricated by Thermacore Inc., was evaluated on Sandias 30 kW Electron Beam Test System equipped with a closed helium flow loop. The module uses short circumferential flow paths to minimize pressure drops and pumping requirements while achieving optimal thermal performance by providing a very large effective surface area. The module was tested under both uniform and non-uniform heat loads to assess the effects of mass flow instabilities. It survived a maximum absorbed heat flux of 29.5 MW/m 2 on a 2-cm 2 area. Results on the power sharing between the two channels is presented and compared with that of a previous design. These experimental results coupled with appropriate modeling provide insight on flow instabilities in multi-channel, helium-cooled heat exchangers.

High heat flux tests on heat sinks armored with tungsten rods

Abstract This paper presents the results of high heat flux tests at Sandia National Laboratories of mock-ups armored with W rods embedded in water-cooled copper-alloy heat sinks. The major result is the excellent performance of these mock-ups in thermal response tests at up to 30 MW/m2 and in thermal cycling tests of 500 cycles (10 s on, 10 s off) at ∼25 MW/m2. Melting of rod tips and the subsequent ‘self-repair’ of a degraded thermal bond were seen. Issues for further development in both testing and manufacturing of the mock-ups were identified. Evaluation of the surface heat flux was found to be somewhat problematic and this issue is discussed.

Prediction of Critical Heat Flux in Water-Cooled Plasma Facing Components Using Computational Fluid Dynamics

Abstract Several commercial computational fluid dynamics (CFD) codes now have the capability to analyze Eulerian two-phase flow using the Rohsenow nucleate boiling model. Analysis of boiling due to one-sided heating in plasma facing components (pfcs) is now receiving attention during the design of water-cooled first wall panels for ITER that may encounter heat fluxes as high as 5 MW/m2. Empirical thermalhydraulic design correlations developed for long fission reactor channels are not reliable when applied to pfcs because fully developed flow conditions seldom exist. Star-CCM+ is one of the commercial CFD codes that can model two-phase flows. Like others, it implements the RPI model for nucleate boiling, but it also seamlessly transitions to a volume-of-fluid model for film boiling. By benchmarking the results of our 3d models against recent experiments on critical heat flux for both smooth rectangular channels and hypervapotrons, we determined the six unique input parameters that accurately characterize the boiling physics for ITER flow conditions under a wide range of absorbed heat flux. We can now exploit this capability to predict the onset of critical heat flux in these components. In addition, the results clearly illustrate the production and transport of vapor and its effect on heat transfer in pfcs from nucleate boiling through transition to film boiling. This article describes the boiling physics implemented in CCM+ and compares the computational results to the benchmark experiments carried out independently in the United States and Russia. Temperature distributions agreed to within 10 °C for a wide range of heat fluxes from 3 MW/m2 to 10 MW/m2 and flow velocities from 1 m/s to 10 m/s in these devices. Although the analysis is incapable of capturing the stochastic nature of critical heat flux (i.e., time and location may depend on a local materials defect or turbulence phenomenon), it is highly reliable in determining the heat flux where boiling instabilities begin to dominate. Beyond this threshold, higher heat fluxes lead to the boiling crisis and eventual burnout. This predictive capability is essential in determining the critical heat flux margin for the design of complex 3d components.

Ultra Low Pressure-Drop Helium-Cooled Porous-Tungsten PFC

Abstract A new class of helium-cooled high heat-flux plasma facing heat exchanger (HX) concept is presented. These unique “Foam-In-Tube” HX concepts are composed of a thin tungsten shell integrally bonded to an open-cell tungsten foam core. High heat flux tests show maximum heat loads of 22.4 MW/m2 using 4 MPa helium at a flow rate of 27 g/s. Based on these impressive performance results, a unique and scalable heat exchanger channel with ultra-low pressure drop through the porous foam is presented. The primary advantage of the new concept is that pressure drop through the porous media and structure temperatures are nearly independent of HX tube length. The concept is modular in design and can be combined to meet divertor size requirements. From a manufacturing and reliability point of view, the advantage of the proposed concept is that it minimizes the need for joining to other functional materials.

Comparison of electron beam test facilities for testing of high heat flux components

Abstract In the last few years, electron beam facilities for the testing of high heat flux components have been erected in Europe, Japan, Russia and in the USA. In principle all the facilities are comparable, but some machine parameters are quite different. These differences include electron beam operation (beam generation, beam diameter, sweeping mode), as well as the temperature measurement devices, calibration techniques and the definition of absorbed power densities. In order to assess the influence of these machine parameters and techniques on the results of high heat flux experiments, a round robin test has been performed in five facilities. In these tests, actively cooled CFC monoblock mock-ups were heated by electron beams using target power densities up to 15 MW/m 2 . Mock-up temperatures and their distribution, measured by different methods (IR camera, pyrometer, thermocouples), have been used as criteria for comparison. The evaluation of data from the different facilities shows good agreement for identical target loading conditions.

Fabrication and high heat flux testing of plasma sprayed beryllium ITER first wall mock-ups

Abstract Plasma-sprayed beryllium ITER first wall mock-ups have survived 3000 thermal fatigue cycles at 1 MW/m 2 without damage during testing at the Plasma Materials Test Facility at Sandia National Laboratory in New Mexico. This heat flux level is twice the expected design heat flux for ITER first wall modules. Plasma sprayed beryllium mock-ups were vacuum plasma sprayed at the Los Alamos National Laboratorys Beryllium Atomization and Thermal Spray Facility. Results will be reported on the fabrication, high heat flux testing and post-mortem analysis of two beryllium plasma sprayed mock-ups (1) beryllium plasma sprayed directly on a CuNiBe heat sink and (2) beryllium plasma sprayed on a compliant layer of aluminum which was explosion-bonded to a CuCrZr copper heat sink. The high heat flux tests utilized the 30 kW Electron Beam Test System at a Sandia National Laboratory.