P. Komarek

Durham University, Durham, England, United Kingdom

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Publications (44)33.39 Total impact

  • P. Komarek
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    ABSTRACT: Superconducting magnets are stringent for fusion reactors with magnetic confinement to provide an economic energy balance. Large-scale development programmes have been executed worldwide to achieve in time the needed technology. The ultimate result of this effort, so far, is the ITER magnet system with the most sophisticated LTS. However, if one thinks about the design of a fusion DEMO and later reactors, the option of HTS must be considered seriously in view of the potential advantages of these conductors concerning higher operation temperature, temperature margin, high field properties and cryogenic power saving. Extrapolating from the long period needed for the ITER conductor development, it is time now to start with HTS development for fusion reactors to be able to decide their applicability. It is still a long way for the HTS to become comparable with LTS, however, continuous progress can be seen. Beside the large confinement magnets, HTS will also be of advantage for current leads, bus bars and gyrotron magnets. The state of the art of HTS is such, that already now such components can be constructed with HTS.
    Fusion Engineering and Design 11/2006; DOI:10.1016/j.fusengdes.2006.07.040 · 1.15 Impact Factor
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    ABSTRACT: The use of high temperature superconductor (HTS) materials in future fusion machines could increase the efficiency drastically, but strong boundary conditions exist. To outline the prospects, challenges and problems, first the benefit of using HTS materials is estimated considering the saving in cryogenic power. Next, it is demonstrated that industrial available HTS materials can be used for fusion today. For this purpose, we give a short summary of results that have been obtained from an ITER conform 70 kA HTS current lead that was designed, built and tested by the Forschungszentrum Karlsruhe and the CRPP Villigen in the frame of the European Fusion Technology Programme and in cooperation with industry. This current lead consists of an HTS part that covered the temperature range from 4.5 to 70 K and a conventional part, making the connection to room temperature. Because the HTS part had no ohmic losses and poor thermal conduction, the refrigerator power necessary for cooling the current lead was reduced drastically. The saving factor could be calculated to be 5.4 at zero current and 3.7 at 68 kA. The current lead could even be operated at 80 kA and with respect to safety criteria of ITER, a complete loss of He flow was simulated showing that the HTS current lead could hold a current of 68 kA for 6 min without active cooling. These results demonstrate that today existing HTS materials can be used in ITER for current leads or bus bar systems.For fusion machines beyond ITER, the development of an HTS fusion conductor would be the key to operate the complete magnet system at higher temperatures. The option of developing fusion conductors based on Bi-2223 and YBCO are briefly discussed. For a success of such conductors, the AC loss optimisation is crucial.
    Fusion Engineering and Design 11/2005; DOI:10.1016/j.fusengdes.2005.06.198 · 1.15 Impact Factor
  • P. Komarek
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    ABSTRACT: To ensure the availability of the ITER magnets an extensive development programme has been executed to demonstrate that the technology is available on industrial scale. This programme included two major projects, firstly the development and testing of a central solenoid model coil (CSMC) with additional testing of high field inserts and secondly the development and testing of a toroidal field model coil (TFMC), both supported by investigations on conductors, structural material and critical components.The model coils have been successfully developed, fabricated and tested in dedicated test facilities at JAERI/Naka (CSMC) and Forschungszentrum Karlsruhe (TFMC), respectively. The tests confirmed that the design criteria are met by the model coils and the ITER coils can be built on this basis. Of course, the tests showed also in which areas improvements should be implemented. One concerns the Nb3Sn conductor. To guarantee the defined margin of the current sharing temperature, advanced conductor strands with higher critical current density than those used in the model coils have to be used and are available now. The codes for mechanical behaviour, ac loss predictions and thermohydraulic behaviour gave rather good agreement with the measurements, but can still be improved. Quality assurance issues concern high voltage properties and leak tightness.
    Fusion Engineering and Design 11/2005; DOI:10.1016/j.fusengdes.2005.08.016 · 1.15 Impact Factor
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    ABSTRACT: The ITER toroidal field model coil (TFMC) was designed, constructed and tested by the European Home Team in the framework of the ITER research and development program of the Engineering Design Activities (EDA). The project was performed under the leadership of European Fusion Development Activity/Close Support Unit (EFDA/CSU), Ciarching. in collaboration with the European superconductor laboratories and the European industry. The TFMC wits developed and constructed in collaboration with the European industry consortium (AGAN) and Europa Metalli LMI supplied the conductor, The TFMC was tested in the test phase I as single coil and in phase 11 in the background field of the EURATOM LCT coil in the TOSKA facility of the Forschungszentrum Karlsruhe. In phase 1, the TFMC achieved an ITER TF coil relevant current of about 80kA and further representative test results before the end of the EDA. In the more complex test phase [I. the coil was exposed to ITER TF coil relevant mechanical stresses in the winding pack and case. The tests confirmed that engineering design principles and manufacturing procedures are sound and suitable for the ITER TF full size coils. The electromagnetic. thermo hydraulic and mechanical operation parameters agree well with predictions. The achieved Lorentz force on the conductor was about 800 kN/m. That has been equivalent to the Lorentz forces in ITER TF coils. (c) 2005 Elsevier B.V. All fights reserved.
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    ABSTRACT: The development of superconducting systems for electric power is driven by the promise of improved efficiency, smaller size, and reduced weight as compared to existing technologies and by the possibility of new applications. Superconducting power components can also contribute to improved power quality and increased system reliability. This paper addresses historical developments and technology status of four superconducting power applications: cables, superconducting magnetic energy storage (SMES), fault-current limiters, and transformers. Today, SMES is the only fully functional superconducting system and it has seen only limited use at grid power levels. A few model or demonstration units exist for each of the other three applications. Superconductivity faces several hurdles on the path to widespread use. Perhaps the most important is the need for operating voltages of 100 kV or more. Though progress in this and other areas has been rapid, considerable development is needed before superconducting devices perform reliably in the utility environment. As a result, today, most initial installations are aimed at niche applications and will be installed where space is limited, where power demands are increasing over existing corridors, and/or where initial development costs can be offset by enhanced power grid performance.
    Proceedings of the IEEE 11/2004; DOI:10.1109/JPROC.2004.833674 · 5.47 Impact Factor
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    ABSTRACT: The tests of the toroidal field model coil (TFMC) were completed in 2002 in the TOSKA facility of Forschungszentrum Karlsruhe, Germany. Operation reached a combined 80 kA in the TFMC and 16 kA in the LCT coil, resulting in a peak electromechanical load very close to that expected in the full-size ITER TF coils (800 kN/m). Here we concentrate on the measurements of the current sharing temperature T<sub>cs</sub> of the TFMC conductor, possibly the highlight of the whole test campaign. These tests were performed by increasing in steps the helium inlet temperature T<sub>in</sub> in double pancake DP1, resulting in an increasing normal voltage V across the DP1.1 and DP1.2 conductors, and were repeated for several combinations of currents in the TFMC and in the LCT coil. The analysis of the V - T<sub>in</sub> characteristic by means of the M&M code allows to self-consistently deriving an estimate of T<sub>cs</sub>, as well as an indirect assessment of the "average" strain state in the conductor. The TFMC isolated strand has also been very recently characterized at different applied uniaxial strain, and preliminary results indicate a stronger reduction of carrying capacity compared to the extrapolation from Summers scaling used in the analysis so far. As a consequence, the performance of the TFMC conductor, as preliminarily re-evaluated here, appears more in line with the strand performance than in previous analysis, although a BI-dependent "degradation" is still present.
    IEEE Transactions on Applied Superconductivity 07/2004; DOI:10.1109/TASC.2004.830682 · 1.32 Impact Factor
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    ABSTRACT: The ITER Toroidal Field Model Coil (TFMC) was designed and manufactured by the European Home Team in collaboration with European industry. The test in the TOSKA facility of the Forschungszentrum Karlsruhe was successfully performed in 2001 and 2002 and has confirmed that the used design and construction principles are applicable for the ITER TF coils. The TFMC was tested up to the rated current of 80 kA as a single coil and in the background field of the EURATOM LCT coil in order to achieve ITER TF coil relevant stress levels. For the operation of the TFMC and LCT coils, special developed forced-flow-cooled current leads were used. Both coils with a total weight of 108 t were forced-flow-cooled with supercritical He at 4.5 K in a secondary cooling loop connected to the 2 kW refrigerator. However, for currents above 11.4 kA in the LCT coil, its winding had to be cooled at 3.0 K with a separate refrigerator and cooling system. Details of the process engineering of both cooling systems will be described. The operation experiences during cool down, standby and current operation and recooling after fast discharges or Tcs measurements will be outlined hereafter.
    06/2004;
  • P. Komarek
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    ABSTRACT: Within the ITER engineering design activity (EDA) seven large R&D projects have been executed to demonstrate the feasibility of the ITER construction. The "L-2" project contained the design, construction and testing of a subsize TF coil, the toroidal field model coil (TFMC), and the manufacturing of two full size sections of a TF coil case. The objective of the L-2 project was to develop and demonstrate the superconducting magnet technology to a level allowing the ITER coils to be built and operated with confidence. The TFMC has been designed and constructed by the European ITER Home Team in collaboration with European companies under the leadership of EFDA/CSU Garching. The testing took place in the TOSKA facility of the Forschungszentrum Karlsruhe in the years 2001 and 2002 with detailed evaluations during this time and afterwards until now. The test arrangement consisted of the TFMC together with a background coil of similar size to expose the TFMC to out-of plane loads and centripetal loads up to stress levels comparable to that of the ITER TF coils. As well suited background coil the reinforced EURATOM LCT coil from the IEA project in the 1980s was used. To minimize risk, in a first test phase the TFMC has been tested alone and then in a second phase together with the LCT coil. This paper gives a survey on the results of this second phase which loaded the TFMC up to its limits, achieving the maximum current ever used in a superconducting magnet so far, namely 80 kA.
    Fusion Engineering, 2003. 20th IEEE/NPSS Symposium on; 11/2003
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    ABSTRACT: For the test of the ITER TFMC in the TOSKA facility of the Forschungszentrum Karlsruhe, two 80 kA current leads were designed and manufactured. Based on the experience coming from the performance of the 30 kA forced-flow current leads, the 80 kA leads were designed in a continuous manner. During the TFMC experiment, various optimization runs were performed at 0, 40 and 80 kA. It could be demonstrated that the leads were operated with the designed mass flow rates. Especially, the Nb<sub>3</sub>Sn inserts used in the lower part of the heat exchanger behave as expected. The slightly different mass flow rate of both terminals can be explained by different RRR of the copper of the heat exchanger. The 80 kA current leads display the highest operating sc coil current reported up to now.
    IEEE Transactions on Applied Superconductivity 07/2003; DOI:10.1109/TASC.2003.812946 · 1.32 Impact Factor
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    ABSTRACT: A high temperature superconducting current lead for application in fusion magnets was developed in collaboration of three laboratories. Bi-2223 Ag/Au tapes were selected to be used in the current lead and two industrial fabricated 10 kA modules were extensively tested. Both modules were used in the 20 kA current lead which was successfully tested in continous operation at 20 kA and over short time up to 40 kA. (C) 2001 Elsevier Science B.V. All rights reserved.
    Fusion Engineering and Design 11/2001; DOI:10.1016/S0920-3796(01)00354-4 · 1.15 Impact Factor
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    ABSTRACT: The EURATOM LCT coil, a D shaped NbTi coil, was tested in the TOSKA facility at Forschungszentrum Karlsruhe with pressurized forced flow He II cooling. A suitable cryogenic system based on an existing 1.8 K He II 300 W liquifier was designed for cooling the coil by pressurized forced flow He II circulated by pumps across heat exchangers immersed in a He II bath. Three types of pumps (piston, centrifugal, thermomechanical) were used. The cryogenic system worked well under usual operation and fault conditions. No significant differences were found between He I and He II forced flow cooling. The D-shaped coil was reinforced by stainless steel belts keeping the D shape in single coil operation mode. No backlash was found between the coil and its reinforcement structure in agreement with predictions of the finite element analysis. The coil achieved ∼11 T as predicted from single strand measurements and previous tests in the frame of the Large Coil Programme. The results are an encouraging step in using this technology in large superconducting magnet systems for magnetic confinement in fusion research.
    Fusion Engineering and Design 08/1999; 45(4-45):361-375. DOI:10.1016/S0920-3796(99)00019-8 · 1.15 Impact Factor
  • P Komarek, E Salpietro
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    ABSTRACT: The TF model coil (TFMC) is one of the seven large R&D projects for the ITER EDA. The TFMC will include most of the technical features and manufacturing approaches foreseen for the full size TF coils and it will be tested in a most relevant manner. This testing will take place 1998/99 in the TOSKA facility of FZK, which is upgraded for that purpose in several steps, each one accompanied by own test purposes, qualifying the facility thereby too. The test arrangement will consist of the TFMC in parallel to the existing EURATOM LCT coil. The availability of this LCT coil with full performance has been reconfirmed by operating it in 1996, even with extended performance at 1.8 K. The major parts of the TOSKA facility are a vacuum vessel with nitrogen shield and a useable volume of 4.30 m diameter and 6.6 m height, a cryogenic supply system with a 2 kW and a 500 W helium refrigerator and cold pumps to circulate helium, two electric power supplies for 30 kA and 50 kA, respectively (they can also be operated in parallel to 80 kA), a further power supply for 20 kA, fast dump circuits with power breakers, the required busbar and current leads and a well suited control and data acquisition system. The status of TOSKA is reviewed, giving details of the components and the qualification tests so far, as the poloidal coil experiment POLO and the LCT coil tests at 1.8 K. A further qualification test will be 1997; the tests of the W 7-X stellarator prototype coil in an arrangement similar to the ITER TF model coil testing.
    Fusion Engineering and Design 09/1998; DOI:10.1016/S0920-3796(98)00116-1 · 1.15 Impact Factor
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    ABSTRACT: A superconducting prototype (3 m Ø and nominal current 15 kA) of a poloidal field coil was developed, constructed and tested according to the typical specification of the Tokamak magnet system in collaboration with European industry. In order to withstand fast ramping (≈2 T s−1), plasma disruption (80 T s−1, 10 ms) and plasma control (≈ ±0.05 T, 10 Hz) in the superconducting state a low loss conductor and a well developed high voltage technique at 4 K were indispensable attributes of the design. The sc cable of the conductor consists of mixed matrix strands (NbTi in Cu/CuNi) cooled by a dual cooling system (two phase forced flow helium for heat removal at constant temperature, and stagnant helium for good transient stability). The sc cable is inclosed in a thick walled stainless steel jacket. All coil components are designed and constructed according to the rules of the high voltage technique. The coil was tested in the TOSKA facility at FZK-Karlsruhe. For generating the typical positive and negative magnetic field transients a special high power switching circuit of a discharge power up to 700 MW (30 kA, 23 kV) was constructed and used. The coil was investigated considering its superconducting, electromagnetic and mechanical properties. All results were conclusive in agreement with calculations and fulfil the specifications. A highlight is that for this conductor type the stability against transient field changes is only determined by the expected stability margin or the critical current boundary no other limitations of the ramp rate were observed.
    Fusion Engineering and Design 05/1997; 36(2-3-36):227-250. DOI:10.1016/S0920-3796(97)00005-7 · 1.15 Impact Factor
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    ABSTRACT: The aims of the toroidal field (TF) model coil are to test the manufacturing feasibility of the ITER TF magnet concepts, to assess the reliability of the fully integrated system by dedicated testing and to qualify the quality assurance and the testing methods. The 3.8 m long racetrack shaped TFMC will be tested in the TOSKA facility at Karlsruhe in a configuration using the Euratom LCT Coil to provide in-plane and out-of-plane loads, relevant to the ITER coils
    IEEE Transactions on Magnetics 08/1996; DOI:10.1109/20.508616 · 1.21 Impact Factor
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    ABSTRACT: A prototype tokamak poloidal field coil of 3 m diameter was successfully tested. The coil was designed to withstand typical field transients like fast ramping, plasma controlling and plasma disruptions. In order to achieve these requirements, a low loss conductor was developed. The coil with its components was designed according to rules of high voltage technology. A realistic test of the magnet requires at 23 kV a current of 15 kA, that means a peak power of 345 MW. Therefore, a fast switching circuit was used combined with an additional current lead at the middle of the magnet winding. This allowed the performance of fast field changes with increasing or decreasing field. In one operation mode, the coil was discharged into a dump resistor. In another operation mode, only one coil half is discharged which generated a field increase in the second short circuited half of the coil. Discharging time constants of 1 ms and local field sweep rates of about 240 T/s were obtained without quench. A reproducible and sharp quench boundary was found. The results showed clearly that ramp rate limitations are only given by the expected stability limit or the load line boundary
    IEEE Transactions on Magnetics 08/1996; DOI:10.1109/20.508615 · 1.21 Impact Factor
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    ABSTRACT: An efficient finite element formulation of the eddy current field in thin conducting sheets has been developed coupled to a description of the magnetic field by a gauged magnetic vector potential. The numerical problems associated with the thin conductors have thereby been eliminated. Results of computations on a model of the TOSKA Upgrade test facility are presented. The aluminium radiation shields are not in danger; the stresses acting upon them are negligible. This is due to their being located relatively far from the superconducting coil. Only parts of the supply lines near the connecting flange of the vacuum vessel experience considerable stress. Their reinforcement under rapid changes may be necessary. The vacuum vessel is not endangered, since its wall of 20 mm thickness easily withstands the stresses due to the induced eddy currents. Normal pressures of up to about 50 t/m<sup>2</sup> occur at the LN2 shield. This makes modifications of its structure necessary
    IEEE Transactions on Magnetics 04/1992; DOI:10.1109/20.123983 · 1.21 Impact Factor
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    ABSTRACT: A superconducting cable for a fusion magnet has to be reinforced by a stainless steel jacket which envelops the cable. The jackets are fabricated by extruded and butt welded steel sections or by composing the jacket from drawn sections with longitudinal welds. Fatigue life and fatigue crack growth rates (FCGRs) were investigated at 10 K for embedded and surface cracks for the candidate material 316LN. Embedded cracks are more severe than surface cracks. The FCGR of weldments is low compared to their base metals, and the residual stresses are responsible for these results. In the range 20 K-7 K, the FCGR of surface cracks is lower than that of cracks propagating from a buried penny-shaped crack
    IEEE Transactions on Magnetics 02/1992; DOI:10.1109/20.119853 · 1.21 Impact Factor
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    ABSTRACT: A forced flow Nb<sub>3</sub>Sn subsize conductor for the 12-T Next European Torus (NET) toroidal field coils fabricated by the react-and-wind process was developed. It consists of a flat core housing the Nb<sub>3</sub>Sn cable soldered between two stabilizing copper units which are surrounded by a stainless steel jacket. The jacket was tightly drawn onto the already reacted cable. Critical current, I <sub>c </sub>, vs. both magnetic field, B , and axial strain, ∈ <sub>a</sub>, measurements show that this delicate procedure has not adversely affected the I <sub>c</sub> vs. ( B ,∈<sub>a</sub>) characteristic. At B =12 T the I <sub>c</sub> vs. ∈<sub>a</sub> curve exhibits a maximum of I <sub>c</sub>=5.25 kA at a prestrain of ∈<sub>a</sub>=0.36% which degrades to I <sub>c</sub>=3.93 kA at ∈<sub>a</sub>=0. This behavior is not influenced by 10<sup>3 </sup> strain cycles between 0.1 and 0.36% strain. The results have demonstrated both the mechanical integrity and the predicted properties of react-and-wind conductors
    IEEE Transactions on Magnetics 04/1991; DOI:10.1109/20.133574 · 1.21 Impact Factor
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    ABSTRACT: The aim of the Large Coil Task (LCT) was to demonstrate the reliable operation of large superconducting toroidal field (TF) coils and to prove the design principles and fabrication techniques to be applied for the magnets in a tokamak experimental power reactor. This has been achieved by an outstanding international development effort during more than ten years of cooperation within an IEA agreement. Parties were the US DOE, EURATOM, JAERI and the Swiss government. Six different D-shaped test coils were separately designed, developed and constructed by the LCT participants, then extensively tested together in a compact toroidal array. The ORNL acted for DOE as the LCT operating agent, building and operating the required test facility. The US also provided three test coils; the other three participants one coil each. Detailed information on coil design and manufacture and all test data were shared among the LCT participants. After facility shakedown operations and preliminary coil tests, the full six-coil array tests were carried out in a continuous period from the beginning of 1986 until September 1987. Beside the originally planned tests to reach an 8 T design point performance, the tests went well beyond this goal, reaching 9 T peak field in each coil. The experiments also delineated the limits of operability and demonstrated the coil safety under abnormal conditions. For fusion application the transient a.c. field behaviour in the coils was also of great interest. Three of the coils have been tested in this respect and showed excellent performance, with loss values in agreement with the theoretical predictions. At the time of International Experimental Reactor (ITER) activities, it might be worthwhile to mention that LCT demonstrated an effective multinational collaboration in an advanced technology project, involving large scale hardware produced in several countries then assembled and operated as a tightly integrated system.
    Cryogenics 09/1989; DOI:10.1016/0011-2275(89)90199-9 · 0.94 Impact Factor

Publication Stats

286 Citations
33.39 Total Impact Points

Institutions

  • 2004
    • Durham University
      Durham, England, United Kingdom
  • 1985–2003
    • Karlsruhe Institute of Technology
      • Institut für Technische Physik
      Karlsruhe, Baden-Wuerttemberg, Germany
    • James Madison University
      Harrisonburg, Virginia, United States
  • 1988
    • Oak Ridge National Laboratory
      Oak Ridge, Florida, United States