B. Schunke

Princeton University, Princeton, New Jersey, United States

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Publications (88)126.47 Total impact

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    ABSTRACT: The neutral beam (NB) system for ITER is composed of two heating neutral beam injectors (HNBs) and a diagnostic neutral beam injector (DNB). A third HNB can be installed as a future up-grade. This paper will present the design development of the components between the injectors and the tokamak; the so-called ‘front end components’: the drift duct consists of the NB bellows and the drift duct liner, the vacuum vessel pressure suppression system box (VVPSS box), the absolute valve, and the fast shutter. These components represent the key links between the ITER tokamak and the vessels of the NB injectors. The design of these components is demanding due to the different loads that these components will have to stand.
    Fusion Engineering and Design 10/2013; 88(9-10):2110-2114. DOI:10.1016/j.fusengdes.2013.02.131 · 1.15 Impact Factor
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    ABSTRACT: The ITER neutral beam (NB) injectors are used for heating and diagnostics operations. There are 4 injectors in total, 3 heating neutral beam injectors (HNBs) and one diagnostic neutral beam injector (DNB). Two HNBs and the DNB will start injection into ITER during the hydrogen/helium phase of ITER operations. A third HNB is considered as an upgrade to the ITER heating systems, and the impact of the later installation and use of that injector have to be taken into account when considering the installation and assembly of the whole NB system. It is assumed that if a third HNB is to be installed, it will be installed before the nuclear phase of the ITER project.
    Fusion Engineering and Design 10/2013; 88(9-10):2029-2032. DOI:10.1016/j.fusengdes.2013.03.010 · 1.15 Impact Factor
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    ABSTRACT: CEA has undertaken tests to study the resilience of copper electrodes in vacuum against energetic high-voltage breakdowns using external capacitors to provide the energy. Earlier tests succeeded in dissipating a maximum of 150 J in a 30 mm gap, limited by the equivalent series resistance (ESR) in the external capacitors. Using new ones with an ESR that is a factor of 10 lower it was unsuccessfully tried to produce breakdowns at 200 kV over the 85 mm gap, despite the use of a UV flash lamp and a “field enhancement ring” (FER) that locally increased the electric field on the cathode by 50%. Consequently, the breakdowns had to be produced by raising the voltage to 300–350 kV while maintaining the gap at 85 mm. During these tests, single breakdowns dissipated up to 1140 J in the 85 mm vacuum gap. Inspection of the electrodes revealed that substantial amounts of copper appear have been evaporated from the anode and deposited on to the cathode. Also electrode deconditioning occurred.
    Fusion Engineering and Design 10/2013; 88(6-8):891-894. DOI:10.1016/j.fusengdes.2012.12.012 · 1.15 Impact Factor
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    ABSTRACT: The 100 kV negative hydrogen ion source based Diagnostic Neutral Beam (DNB) injector is part of the Indian procurement for ITER. The DNB is designed to deliver an 18–20 A hydrogen neutral beam to the ITER plasma. The exit scraper (ES) defines the profile of the beam after it leaves the calorimeter, the last Beam Line Components (BLCs) during operation. BTR1 and PDP1 codes are used to obtain the optimum entry and exit opening dimensions of the ES, so as to deliver the maximum power to the tokamak plasma, and to limit the power to downstream components due to beam interception, to protect front end components from the interception of re-ionized devious particles. Each horizontal heat transfer (HT) panel of ES receives a total power of 85 kW for each symmetric 10 mrad beam (worst case scenario) with peak power density of 1.17 MW/m2. The thermo hydraulic design is carried out to withstand the heat flux due to the beam interception on the ES. The mechanical design of the ES is carried out by considering the spatial constraints in the DNB system and the remote handling (RH) system. In the present design, the copper heat transfer panels are directly bolted to the structural member. Protection shall be incorporated to ensure that there is no exposure of the bolt head. The thermo-mechanical analysis has been carried out for the normal, off-normal (accidental/worst-case) events of the ITER DNB. During normal operation, the maximum possible von Mises stress of the order of 62.4 MPa can be expected on the bottom heat transfer panel, and the corresponding deflection of the structure is of the order of 0.62 mm. All the stress checks were carried out with respect to the Structural Design Criteria for In-vessel Components (SDC-IC) criterion. The paper shall present the ES design highlighting its salient features.
    SOFT 2012; 03/2013
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    ABSTRACT: Positioning of Cesium (Cs) oven modules in the complex interface dominated space envelope of a negative ion source such as Diagnostic Neutral Beam (DNB) source for ITER is a challenge not only for the designer of the ion source, but also that of remote handling. A more user friendly design of the Cs delivery could emerge from the consideration of a possibility of injecting the Cs from an oven located outside the vacuum envelope of the ion source, thereby ensuring an ease of Cs refilling and oven maintenance. The design of such a delivery system involves long transmission path of lengths ~4 m, from ambient to vacuum. System design involves incorporation of a low loss transmission tube enveloped by highly reflective inner surface pipe to reduce the heat losses and therefore heating of the nearby systems. A combination of all metallic valves operated at high temperatures has been incorporated in such a way that the Cs refilling or oven maintenance can be done without breaking the ion source vacuum. Removable joints in the oven heating elements are provided at specific locations to remove the Cs oven for ion source maintenance. Experimental data on Cs transmission over such a long length, required for an effective design of a co-axial transmission, is not presently available. However, an experiment has been carried out in ITER-India making measurements of Cs distribution in coaxial transmission of a length of more than 5 m. These experiments incorporate an additional feature of multiple nozzle distributor based Cs delivery into the ion source which might help in reducing the need of multiple Cs ovens in large ion sources like ITER. The Cs flux from the oven is measured by surface ionization detector (SID). The angular distribution of the Cs flux is measured by a movable SID in linear direction and has been found in good agreement with the calculations. The Cs inventory in the oven reservoir was measured by electrical resistivity measurements methods. The paper proposes to present the measurement results and also proposes a possible configuration of the Cs oven for ITER DNB ion source.
    02/2013; DOI:10.1063/1.4792787
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    ABSTRACT: Calorimeter for Diagnostic Neutral Beam (DNB) consists of two movable panels, forming V-shaped configuration with an included angle of 45 Deg. w.r.t. neutral beam axis. Panel's closing and opening motion will be achieved through Calorimeter Motion Mechanism (CMM). The DNB calorimeter design has evolved after analyzing several design proposals serving the basic functionality of the component, but these designs had compatibility issues for the ITER Remote Handling (RH) scenarios, operational maintenance, ITER Vacuum Handbook (IVH) compliance for hydraulic integration, design of flexible elements for large displacements, neutron irradiation susceptibility and joint materials for Vacuum Quality Class 1 (VQC1) application. In the design concept the movement mechanism and hydraulic line flexible element have been deliberately kept outside the DNB vessel, hence there is no sliding and complicated links inside the vessel which cannot be maintained through available RH tools. The paper proposes to present state of the art design of the ITER-DNB Calorimeter.
    Fusion Engineering (SOFE), 2013 IEEE 25th Symposium on; 01/2013
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    ABSTRACT: The ITER equatorial port visible∕infrared wide angle viewing system concept is developed from the measurement requirements. The proposed solution situates 4 viewing systems in the equatorial ports 3, 9, 12, and 17 with 4 views each (looking at the upper target, the inner divertor, and tangentially left and right). This gives sufficient coverage. The spatial resolution of the divertor system is 2 times higher than the other views. For compensation of vacuum-vessel movements, an optical hinge concept is proposed. Compactness and low neutron streaming is achieved by orienting port plug doglegs horizontally. Calibration methods, risks, and R&D topics are outlined.
    The Review of scientific instruments 10/2012; 83(10):10E520. DOI:10.1063/1.4734487 · 1.58 Impact Factor
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    ABSTRACT: The ITER fusion device is intended to demonstrate the viability of magnetically confined deuterium-tritium plasma as an energy source. One of the principal methods of heating and driving current in the plasma will be energetic beams of neutral atoms of D° at up to 1 MeV and of H° at up to 870 KeV, with a total injectable neutral beam power after transit through the neutralizer and downstream beamline elements of 16.5 MW from each ion source for pulse lengths of up to 3600 seconds. These requirements far exceed those of any previous positive or negative ion source, thus spawning a substantial development program to ensure that they will be met with a robustly reliable system. The ion source will consist of a large plasma expansion region fed by 8 RF driver units, and will be cesiated to enhance surface production of negative ions, followed by a multi-aperture multi-grid extractor and electrostatic accelerator. The plasma portion of the source is derived from a line of RF sources developed at IPP Garching,1 and the extractor/accelerator from development work at JAEA2, with the integrated design of the ITER source being done at Consorzio RFX. Unlike the first generation of high power negative hydrogen ion isotope sources, the ITER source will have the major advantage of a succession of progressively more comprehensive test facilities, culminating in a full power and pulse length test bed at Consorzio RFX. This talk will discuss the major beamline components, including the ion source and accelerator, the neutralizer cell that converts a portion of the negative ions to neutrals, the residual ion deflection system, and the tokamak field compensation system. Some remaining physics and engineering issues, along with their expected resolution, will be discussed, as well as the development and testing strategy.
    Plasma Science (ICOPS), 2012 Abstracts IEEE International Conference on; 01/2012
  • Lennart Svensson · Ronald S. Hemsworth · Beatrix Schunke
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    ABSTRACT: The ITER Neutral Beam (NB) System consists of a 100 keV diagnostic injector (DNB) and two 1 MeV heating and current drive injectors (HNB). Injection will be done in the different phases of the experimental programme, ranging from low activation operation with hydrogen and helium plasmas, deuterium plasmas, and finally full performance operation with deuterium–tritium mixtures. The injectors will be operated in a harsh environment with high levels of neutron and gamma radiation from the torus. A high stray magnetic field from the tokamak is reduced to acceptable levels for the region up-stream of the residual ion dump (RID) by a combination of a passive magnetic shield and active correction and compensation coils (ACCC). This broad range of requirements imposes a high flexibility, a high availability and reliability of the instrumentation and on the diagnostics that will be used.
    Fusion Engineering and Design 10/2011; DOI:10.1016/j.fusengdes.2011.01.140 · 1.15 Impact Factor
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    ABSTRACT: The 100kV bushing of the Diagnostic Neutral Beam (DNB) injector is a cylindrical feedthrough which forms the interface between the gas insulated transmission line and the torus primary vacuum and provides all necessary services to the beam source. All conventions for safety, voltage holding requirement, vacuum compatibility and the choice of materials have been addressed in the design. Finite Element Analyses (FEAs) for the electrostatic and the structural configuration is carried out to validate the design of the High Voltage Bushing (HVB). Several iterations and optimizations of the stress shields are carried out to meet electrostatic criteria, especially at the triple point (the ceramic, metal and vacuum joint), which is critical for good voltage holding. Structural analyses is carried out to assess the stress distribution in the fiber reinforced plastic insulator, the alumina insulator and the integrated HVB for different load cases like operational orientation (horizontal), normal operation and accidental case. Design is further validated for seismic conditions for Seismic Loading-2 (SL-2).
    Fusion Engineering and Design 10/2011; 86(6):892-895. DOI:10.1016/j.fusengdes.2011.04.041 · 1.15 Impact Factor
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    ABSTRACT: The beam source of neutral beam heating∕current drive system for ITER is needed to accelerate the negative ion beam of 40A with D− at 1 MeV for 3600 sec. In order to realize the beam source, design and R&D works are being developed in many institutions under the coordination of ITER organization. The development of the key issues of the ion source including source plasma uniformity, suppression of co‐extracted electron in D beam operation and also after the long beam duration time of over a few 100 sec, is progressed mainly in IPP with the facilities of BATMAN, MANITU and RADI. In the near future, ELISE, that will be tested the half size of the ITER ion source, will start the operation in 2011, and then SPIDER, which demonstrates negative ion production and extraction with the same size and same structure as the ITER ion source, will start the operation in 2014 as part of the NBTF. The development of the accelerator is progressed mainly in JAEA with the MeV test facility, and also the computer simulation of beam optics also developed in JAEA, CEA and RFX. The full ITER heating and current drive beam performance will be demonstrated in MITICA, which will start operation in 2016 as part of the NBTF.
    09/2011; 1390(1):545-554. DOI:10.1063/1.3637426
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    ABSTRACT: The 100 kV bushing is one of the most important and technologically challenging Safety Important Class (SIC) components of the Diagnostic Neutral Beam (DNB) injector of ITER. It forms interface between gas insulated electrical transmission line and torus primary vacuum and acts as a vacuum feedthrough of ITER. Design optimization has been carried out to meet the electric and structural requirements based on its classification. Unlike HNB bushing, single stage bushing is designed to provide 100 kV isolation. Finite Element Analysis (FEA) based optimization has been carried out for electrostatic and structural analysis. Manufacturing assembly sequence is studied and presented in this paper. However validation of the same is foreseen from manufacturer.
    09/2011; 1390(1):555-566. DOI:10.1063/1.3637427
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    ABSTRACT: A RAMI (Reliability, Availability, Maintainability, Inspectability) analysis has been performed for the heating (& current drive) neutral beam (HNB) and diagnostic neutral beam (DNB) systems of the ITER device [1-3]. The objective of these analyses is to implement RAMI engineering requirements for design and testing to prepare a reliability-centred plan for commissioning, operation, and maintenance of the system in the framework of technical risk control to support the overall ITER Project. These RAMI requirements will correspond to the RAMI targets for the ITER project and the compensating provisions to reach them as deduced from the necessary actions to decrease the risk level of the function failure modes. The RAMI analyses results have to match with the procurement plan of the systems.
    09/2011; DOI:10.1063/1.3637428
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    SOFT; 01/2011
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    ABSTRACT: Inductively coupled radio frequency (RF) based multi-driver negative ion sources form the basis of the ion sources for Neutral Beam (NB) injectors considered for the fusion devices. Database for such multi-driver operations are limited. A need for supplementing the database is paramount for the development of large size negative ion sources. Under Indian program, experiments initiated with the objective of understanding the physics and technology of multi­ driver coupling. Radio frequency (RF) based negative NB systems have addressed to this requirement in the form of configuring an experimental system consisting of a two driver based source powered by a single 1 MHz, 180 kW RF generator. Plasma of density - 1018 m-3, in a volume of - 0.5 m3 chamber shall be produced in this experiment from which, extraction of- 10 -12 A of negative hydrogen ion current @ 50 kV, is foreseen. Power is launched though an actively cooled coil mounted on each of the driver. Adequate flexibility
    SOFE; 01/2011
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    ABSTRACT: ITER will explore a plasma parameter envelope currently not available in tokamaks. This will require a set of diagnostics that can follow this envelope. To implement these diagnostics in a reliable and robust way requires development of current techniques in many areas to make them applicable to ITER: they need to be operable in the ITER environment and satisfy the physics and engineering requirements. In some cases, the exploitation of new techniques will be required. While much work has been carried out in this area, significant further work remains to bring the system to implementation.
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    ABSTRACT: Since the first feasibility studies of active beam spectroscopy on ITER in 1995 the proposed diagnostic has developed into a well advanced and mature system. Substantial progress has been achieved on the physics side including comprehensive performance studies based on an advanced predictive code, which simulates active and passive features of the expected spectral ranges. The simulation has enabled detailed specifications for an optimized instrumentation and has helped to specify suitable diagnostic neutral beam parameters.Four ITER partners share presently the task of developing a suite of ITER active beam diagnostics, which make use of the two 0.5MeV/amu 18MW heating neutral beams and a dedicated 0.1MeV/amu, 3.6MW diagnostic neutral beam. The IN ITER team is responsible for the DNB development and also for beam physics related aspects of the diagnostic. The RF will be responsible for edge CXRS system covering the outer region of the plasma (1>r/a>0.4) using an equatorial observation port, and the EU will develop the core CXRS system for the very core (0
    Nuclear Instruments and Methods in Physics Research Section A Accelerators Spectrometers Detectors and Associated Equipment 11/2010; 623(2):720-725. DOI:10.1016/j.nima.2010.04.011 · 1.32 Impact Factor
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    ABSTRACT: Following the allocation of the procurement of the diagnostic neutral beam (DNB) to the Indian DA, a series of tasks have been undertaken to first assess the DNB configuration and arrive at an optimal beam-line configuration folding in the gas-feed and vacuum-pumping requirements. Specific emphasis is placed on the thermal, structural, and electrical designs of beam-line components, in order to ensure their compatibility with the criteria specified for ITER in vessel components, i.e., Structural Design Criteria for In-Vessel Components. The detailed assessment of manufacturing technologies and their compatibility with the ITER standards forms an integral part of the design. A common approach to manufacturing for DNB and heating-and-current-drive NB components shall be undertaken through a comprehensive prototyping phase which shall lead to built-to-print specifications. In addition to safety and remote-handling issues, the design also addresses the requirements of interfaces related to other systems such as cryo, hydraulic, pneumatic, vacuum pumping, gas feed, civil, power supplies and transmission, CODAC, etc. The successful delivery of DNB is dependent on two critical R&D aspects: 1) the production of a uniform low-divergence beam from the beam source and 2) a well-controlled transmission through lengths of ~ 22 m. The first shall primarily be a subject of the Ion Source Test Facility-SPIDER [part of NB test facility (MITICA in Padova)]-where India is involved as a collaborator and the Indian test bed, where issues for DNB beam source which were not resolved in the SPIDER would be taken up. The second shall form one of the primary objectives of the Indian test bed to characterize the DNB. This paper presents the progress in DNB from the concept level to an engineered system along with the plans for system integration and an R&D intensive implementation.
    IEEE Transactions on Plasma Science 04/2010; 38(3-38):248 - 253. DOI:10.1109/TPS.2009.2035809 · 0.95 Impact Factor
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    ABSTRACT: The 100-kV negative-hydrogen-ion-source-based diagnostic neutral beam (NB) (DNB) injector, which forms a part of the Indian (IN) procurement package for ITER, targets a delivery of ~18-20 A of neutral hydrogen-atom beam current into the ITER torus for charge exchange resonance spectroscopy diagnostics. Considering stripping losses, a ~70-A negative ion current is required to be extracted from the ion source, which leads to a production of 60 A of accelerated ion beam. Subsequent process of neutralization, electrostatic ion separation, and transport to the duct leads to a large separation between the points of generation of the ion beam to the point of delivery of the NB into the torus (~23 m). This forms one of the most important constraints for the transport of NBs to ITER. The requirements are not only for a stringent control over ion optics, the transport to electrostatic separator, minimum loss of beam due to intercepting elements, low reionization loss, and focusing to control interception losses but also for adequate compensation of residual magnetic fields to overcome magnetic field induced deflections also form important design issues for a reasonable transmission efficiency. Due to multiparameter dependence, it becomes necessary to assess the different scenarios using numerical codes. In the present case, the assessment has been carried out for the DNB using the beam-transport codes PDP, BTR, and the MCGF codes which are developed by the Russian Federation. An optimized configuration of the beamline has been arrived at on the basis of these code-enabled studies. These parameters are the following: listing of the vertical and horizontal focal lengths as 20.6 m, a spacing between ground grid and neutralizer of 1 m, and positioning of residual-ion dump at a distance of 0.75 m from the neutralizer exit. Further, optimizing the gas feed to the source and neutralizer leads to a final transmission of ~35% of the extracted beam power to the torus. This paper shall - - present the methodology, the issues concerned, and the final configuration which forms the basis for the present engineering.
    IEEE Transactions on Plasma Science 04/2010; 38(3-38):242 - 247. DOI:10.1109/TPS.2009.2035623 · 0.95 Impact Factor
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    ABSTRACT: The 100 kV negative hydrogen ion source based Diagnostic Neutral Beam (DNB) injector, which forms a part of the IN Procurement Package for ITER, targets a delivery of ~18-20A of neutral hydrogen atom beam current into the ITER torus for charge exchange resonance spectroscopy (CXRS) diagnostics. Considering stripping losses, ~70A negative ion current is required to be extracted from the ion source, which leads to a production of 60 A of accelerated ion beam. Subsequent process of neutralization, electrostatic ion separation and transport to the duct leads to a large separation between the points of generation of ion beam to the point of delivery of the neutral beam into the Torus (~23 m). This forms one of the most important constraints for the transport of neutral beams to ITER. The requirements are not only for a stringent control over ion optics, the transport to electrostatic separator, minimum loss of beam due to intercepting elements, low reionization loss, focusing to control interception losses, adequate compensation of residual magnetic fields to overcome magnetic field induced deflections also forms important design issues for a reasonable transmission efficiency. Due to the multi parameter dependence, it becomes necessary to assess different scenarios using numerical codes. In the present case the assessment has been carried out for the DNB using the beam transport codes PDP, BTR and the MCGF codes which are developed by the Russian Federation. An optimized configuration of the beam line has been arrived at on the basis of these codes enabled studies. These parameters are: listing of the vertical and horizontal focal lengths as 20.6 m, a spacing between ground grid and neutralizer of 1 m, positioning of RID at a distance of 0.75 m from the neutraliser exit. Further, optimizing the gas feed to the source and neutralizer lead to a final transmission of ~35% of the extracted beam power to the torus. The paper shall present the methodology, the issues concerne- d and the final configuration which forms the basis for the present engineering.
    Fusion Engineering, 2009. SOFE 2009. 23rd IEEE/NPSS Symposium on; 07/2009

Publication Stats

800 Citations
126.47 Total Impact Points

Institutions

  • 2012
    • Princeton University
      • Princeton Plasma Physics Laboratory
      Princeton, New Jersey, United States
  • 2005
    • KTH Royal Institute of Technology
      • Department of Physics
      Stockholm, Stockholm, Sweden
  • 2002
    • Cea Leti
      Grenoble, Rhône-Alpes, France
  • 1991–1997
    • University of Toronto
      • Institute for Aerospace Studies
      Toronto, Ontario, Canada
  • 1994
    • University of Milan
      Milano, Lombardy, Italy
    • Université de Montpellier 1
      Montpelhièr, Languedoc-Roussillon, France