C.B. Baxi

Helium-Cooled Refractory Alloys First Wall and Blanket Evaluation

Abstract Under the APEX program the He-cooled system design task is to evaluate and recommend high power density refractory alloy first wall and blanket designs and to recommend and initiate tests to address critical issues. We completed the preliminary design of a helium-cooled, W–5Re alloy, lithium breeder design and the results are reported in this paper. Many areas of the design were assessed, including material selection, helium impurity control, and mechanical, nuclear and thermal hydraulics design, and waste disposal, tritium and safety design. Systems study results show that at a closed cycle gas turbine (CCGT) gross thermal efficiency of 57.5%, a superconducting coil tokamak reactor, with an aspect ratio of 4, and an output power of 2 GWe, can be projected to have a cost of electricity at 54.6 mill/kW h. Critical issues were identified and we plan to continue the design on some of the critical issues during the next phase of the APEX design study.

Review of helium cooling for fusion reactor applications

Abstract Helium cooling has been successfully used for fission reactors in the U.S. and Germany in the past. Helium is an attractive coolant for fusion reactors because it is chemically and neutronically inert and can be used directly for gas turbine cycle power conversion. In addition, as was shown during ITER and other fusion power plant evaluations, it is superior from safety considerations. On the other hand, some researchers are under the impression that use of helium cooling requires high pressure, large pumping power and larger manifold sizes due its low density at atmospheric pressure. In this paper it is shown that these concerns can be eliminated through the use of heat transfer enhancement techniques to reduce the flow, pumping power and pressure requirements. A number of proven heat transfer enhancement techniques such as extended surfaces, swirl tape, roughening, porous media heat exchanger and particulate addition are reviewed. Recent experiments with some of these methods have shown that expected heat fluxes of 10 MW/m 2 in fusion reactors can be removed by helium cooling at a modest pressure of 4 MPa. In this paper designs of divertor heat sinks made from copper, vanadium and tungsten with a peak heat flux of 5–10 MW/m 2 , cooled by helium at a pressure of 4 MPa, are presented.

Thermal hydraulics of water cooled divertors

Abstract Divertors of several new machines, such as JT-60SU, FIRE, SST-1, ITER-RC and KSTAR, are water-cooled. This paper examines critical thermal hydraulic issues associated with the design of such divertors. The flow direction of coolant in the divertor can be either toroidal or poloidal. A quantitative evaluation shows that the poloidal flow direction is preferred, because the flow rate and pumping power are about an order of magnitude smaller compared to the toroidal flow direction. Use of heat transfer enhancement technique leads to a lower water pressure requirement, lower flow rate, lower pressure drop, lower pumping power and lower flow velocity. The swirl tube is the best available method due to a large amount of data and ease of fabrication. The standard critical heat flux (CHF) correlations calculate the CHF value at the surface of the coolant channel, whereas divertor physics studies specify the value of the heat flux incident on the divertor surface. Hence, a finite element thermal analysis of typical divertor geometry was performed. The analysis shows that the ratio of incident heat flux to coolant channel heat flux varies between 1.4 and 1.5 and depends on the heat flux magnitude. The temperatures predicted by the analysis compare very favorably with experimental measurements. The method presented in this paper also correctly predicts incident CHF. A typical axial distribution of heat flux results in about a 20% higher incident CHF compared to a uniform heat flux. Recommended correlations and procedures for water-cooled divertor thermal analyses are presented. The expected peak heat flux of 20 MW/m 2 in tokamak divertors can be accommodated with a flow velocity of about 10 m/s and coolant pressure at an inlet pressure of 3 MPa.

Comparison of swirl tube and hypervapotron for cooling of ITER divertor

The ITER divertor will have a peak steady state heat flux of 5 MW/m/sup 2/ and a heat flux of 15 MW/m/sup 2/ for up to 10 s duration. Cooling will be provided by water at an inlet temperature of 150/spl deg/C and a pressure of 4 MPa. A heat transfer enhancement technique is required in order to achieve a sufficient margin on critical heat flux at a reasonable flow velocity. Hypervapotron (KV) and swirl tube (ST) are under consideration as enhancement methods. There are many fundamental differences between these two devices, such as: (a) The ratio of surface heat flux to coolant channel heat flux, (b) the flow area per unit heat flux area, (c) critical heat flux (CHF) and (d) the pressure drop. This paper presents new CHF correlations for ST and HV concepts and compares them to the available experimental data. The friction factor correlation for ST is well known. A new friction factor correlation for HV based on existing data is presented. A comparison of the two concepts was performed for ITER conditions based on equal heat flux area. The comparison shows that the pumping power required for HV is slightly higher (about 10%) and the incident critical heat flux (ICHF) is slightly lower (8%) for HV compared to ST at similar flow conditions. These differences are small enough and uncertainties in data large enough so that the choice between the two concepts should be based on other considerations such as: (1) cost and ease of fabrication, (2) ease of brazing and (3) volume and reliability of available experimental data: These considerations lead to the conclusion that the choice of concept will depend on the particular application. For ITER, both of these concepts could be used in different areas of the divertor.

Development, installation, and initial operation of DIII-D graphite armor tiles

An upgrade of the DIII-D vacuum vessel protection system has been completed. The ceiling, floor, and inner wall have been armored to enable operation of CIT-relevant double-null diverted plasmas and to enable the use of the inner wall as a limiting surface. The all-graphite tiles replace the earlier partial coverage armor configuration which consisted of a combination of Inconel tiles and graphite brazed to Inconel tiles. A new all-graphite design concept was chosen for cost and reliability reasons. The ten minute duration between plasma discharge required the tiles to be cooled by conduction to the water-cooled vessel wall. Using two and three-dimensional analyses, the tile design was optimized to minimize thermal stresses with uniform thermal loading on the plasma-facing surface. Minimizing the stress around the tile hold-down feature and eliminating stress concentrators were emphasized in the design. The design of the tile fastener system resulted in sufficient hold-down forces for good thermal conductance to the vessel and for securing the tile against eddy current forces. The tiles are made of graphite, and a program to select a suitable grade of graphite was undertaken. Initially, graphites were compared based on published technical data. Graphite samples were then tested for thermal shock capacity in an electron beam test facility at the Sandia National Laboratory (SNLA) in Albuquerque, New Mexico, USA. The expected operational heat load to the tiles is 500 W/cm 2 for 10 s over the entire tile during limiter operation and 1000 W/cm 2 over a 2 cm wide band for 10 s for divertor operation. Comparative stress analysis was performed on the basis of 1000 W/cm 2 uniformly distributed on the front surface for 5 s with 10 min between pulses. Prototype tiles survived beam heating to the limit of the beam system which was 1000 W/cm 2 for 4 s or 1750 W/cm 2 for 1.75 s. The electron beam facility at the Sandia National Laboratory was used to deposite higher heat flux in a 2 cm wide band in order to model divertor heat flux conditions. These tests were successful with heat fluxes up to 2500 W/cm 2 for 5 s. Both the ion beam and the electron beam tests produced surface temperatures up to 3100 °C which resulted in minor surface erosion but no structural cracking. Installation of approximately 1600 armor tiles was accomplished in November/December 1987. Initial operation experience with the new armor system is summarized.

Evaluation of helium cooling for fusion divertors

Abstract The divertors of future fusion reactors will have a power throughput of several hundred megawatts. The peak heat flux on the divertor surface is estimated to be 5–15 MW m−2 at an average heat flux of 2 MW m−2. The divertors have a requirement of both minimum temperature (100°C) and maximum temperature. The minimum temperature is dictated by the requirement to reduce the absorption of plasma, and the maximum temperature is determined by the thermomechanical properties of the plasma facing materials. Coolants that have been considered for fusion reactors are water, liquid metals and helium. Helium cooling has been shown to be very attractive from safety and other considerations. Helium is chemically and neutronically inert and is suitable for power conversion. The challenges associated with helium cooling are (1) manifold sizes, (2) pumping power, and (3) leak prevention. In this paper the first two of the above design issues are addressed. A variety of heat transfer enhancement techniques is considered to demonstrate that the manifold sizes and the pumping power can be reduced to acceptable levels. A helium-cooled divertor module was designed and fabricated by GA for a steady-state heat flux of 10 MW m−2. This module was recently (August 1993) tested at Sandia National Laboratories. At the inlet pressure of 4 MPa, the module was tested at a steady-state heat flux of 10 MW m−2. The pumping power required was less than 1% of the power removed. These results verified the design prediction.

Engineering design of cryocondensation pumps for the DIII-D radiative divertor program

A new double-null, slotted divertor configuration will be installed for the DIII-D Radiative Divertor Program at General Atomics in late 1996. Four cryocondensation pumps, three new and one existing, will be part of this new divertor. The purpose of the pumps is to provide plasma density control and to limit the impurities entering the plasma core by providing pumping at each divertor strike point. The three new pumps are based on the design of the existing pump, installed in 1992 as part of the Advanced Divertor Program. The pump continues to operate successfully. The toroidally continuous pumps vary in lengths from approximately 7 to 12 m depending upon their locations within the vessel. Each pump is independently operated and offers on average 0.7 m/sup 2/ of liquid helium-cooled pumping surface. The tubular pumping surface is surrounded concentrically by nitrogen shields and a particle shield of larger diameters. The nitrogen-cooled shields limit the heat flux on the helium surface. The particle shield limits energetic particles from impacting the helium and nitrogen cooled surfaces, preventing the condensed gases on the pump, primarily water, from being released. The new pumps require geometry modifications to the original design. Therefore, extensive modal and dynamic analyses were performed to determine the behavior of these pumps and their helium and nitrogen feed lines during disruption events. Thermal and fluid analyses were also performed to characterize the helium two-phase flow regime in the pumps and their feedlines. A flow testing program was completed to test the change in geometry of the pump feed lines with respect to helium flow stability. The results were compared to the helium thermal and fluid analyses to verify predicted flow regimes and flow stability.

Design, fabrication, installation and testing of in-vessel control coils for DIII-D

Abstract Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. Recently, these coils have also demonstrated significant benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modelling has shown that substantial performance improvements can be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and have been tested successfully. Installation of a set of twelve internal coils and magnetic sensors in the DIII-D tokamak is to be completed in December 2002. The design requirement for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator.

Design, fabrication, installation, testing and initial results of in-vessel control coils for DIII-D

Since 1995, DIII-D has performed correction of magnetic field imperfections using a set of six external picture frame coils located on the vessel mid-plane. In 2000, these coils also demonstrated benefits when used for feedback of the resistive wall mode, an instability that limits the plasma performance at high beta. Modeling has shown that substantial performance improvements could be achieved by installing new coils inside the vessel and expanding the poloidal coverage above and below the mid-plane. Two prototype internal coils were installed in 2001 and were power tested successfully after several bakes to 350/spl deg/C. A full set of twelve internal coils and related magnetic sensors are now operational in the DIII-D tokamak. The design requirements for the new coil system was to maximize the magnetic field at the plasma edge, operate with a frequency range of dc to 1000 Hz, and fit behind the existing graphite wall tiles. The coil design adopted and installed is a water-cooled hollow copper conductor insulated with polyamide and housed inside a stainless steel tube that forms a vacuum boundary. The coil is rigidly mounted to the inside of the vacuum vessel. The primary challenge in the design of these coils was in joining of both the copper conductor and the stainless tube without overheating the polyamide insulator. Elastic-plastic analysis was used to demonstrate acceptable thermal stresses during baking conditions. Analysis determined the optimum water cooling channel diameter. The coils were tested in high toroidal field to the limit of the power supply of 4.5 kA DC with inductance-limited current for frequencies between 300 Hz and 1000 Hz. Recent results are presented.

Analytical prediction of thermal performance of hypervapotron and its application to ITER

A hypervapotron (HV) is a water cooled device made of high thermal conductivity material such as copper. A surface heat flux of up to 30 MW/m{sup 2} has been achieved in copper hypervapotrans cooled by water at a velocity of 10 m/s and at a pressure of six bar. Hypervapotrons have been used in the past as beam dumps at the Joint European Torus (JET). It is planned to use them for diverter cooling during Mark II upgrade of the JET. Although a large amount of experimental data has been collected on these devices, an analytical performance prediction has not been done before due to the complexity of the heat transfer mechanisms. A method to analytically predict the thermal performance of the hypervapotron is described. The method uses a combination of a number of thermal hydraulic correlations and a finite element analysis. The analytical prediction shows an excellent agreement with experimental results over a wide range of velocities, pressures, subcooling, and geometries. The method was used to predict the performance of hypervapotron made of beryllium. Merits for the use of hypervapotrons for International Thermonuclear Experimental Reactor (ITER) and Tokamak Physics Experiment (TPX) are discussed.