P. Titus - Academia.edu (original) (raw)
Papers by P. Titus
MIT, PSFC is in the process of designing and installing a cryopump in the Alcator C-MOD vessel. S... more MIT, PSFC is in the process of designing and installing a cryopump in the Alcator C-MOD vessel. Similar to that used at D-III-D [2], the cryopump will be used for density and impurity control during h-mode, ITB, and lower hybrid current drive operation on C-MOD. Performance goals are discussed in a companion paper at this conference on testing and installation. The mechanical design and analysis of the pump is presented in this paper. Significant loading includes differential contraction, pressurization due to up to to air fault in the vessel, disruption eddy currents, and halo currents in the tile shelf and supporting structures. The cold component support system is described and simulations of cooldown and pressurization are presented. Eddy currents are computed from a transient electromagnetic ANSYS simulation. This provides toroidal currents which in turn are input to a simulation of the pulse in which the cryopump disruption eddy currents are assumed to occur at any time in the...
2015 IEEE 26th Symposium on Fusion Engineering (SOFE), 2015
The Wendelstein 7-X (W7-X) stellarator is in final stages of commissioning, and will begin operat... more The Wendelstein 7-X (W7-X) stellarator is in final stages of commissioning, and will begin operation in the last half of 2015. In this first phase the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady-state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current may evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A high heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a t...
Fusion Technology, 1996
A new type of magnet design is proposed, where the conductor is composed of conducting pebbles in... more A new type of magnet design is proposed, where the conductor is composed of conducting pebbles in a low-melting temperature conducting matrix. The magnet should have high radiation tolerance. At the end-of-life of the conductor, the pebbles can be circulated out of the magnet after the conducting matrix is molten. Application of this approach to the centerpost in the Low Aspect Ratio Device is discussed. 6 refs., 1 fig.
IEEE/NPSS Symposium on Fusion Engineering, Sep 30, 1991
Mandatory standards for magnet structural design have been proposed. They pertain to all phases o... more Mandatory standards for magnet structural design have been proposed. They pertain to all phases of magnet technology from conceptual design through lifetime performance, and are intended for all magnet types. The basis for development of the standards is presented, together with extracted sections of the structural guidelines document that preceded it, with a few examples to support conservatism in the design of magnet systems.<<ETX>>
Fusion Science and Technology, 2015
The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the co... more The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs. *higher β N operation can allow higher neutron wall load **1 year has been allocated at the end of each phase in addition to the specified maintenance time
Fusion Technology, 1996
The President`s Committee of Advisors on Science and Technology (PCAST) has suggested that a devi... more The President`s Committee of Advisors on Science and Technology (PCAST) has suggested that a device with a mission of ignition and moderate burn time could address the physics of burning plasmas at a lesser cost than ITER with its more comprehensive physics and technology mission. The Department of Energy commissioned a study to explore this PCAST suggestion. This paper describes the results of the engineering portion of the study of this `PCAST Machine;` physics is covered in a companion paper authored by G.H. Neilson, et al; and the costs are covered in a companion paper by R.T. Simmons, et al. Both are published in the proceedings of this conference. The study was undertaken by a team under the direction of Bruce Montgomery that included representatives from MIT, PPPL, ORNL, LLNL, GA, Northrup-Grumman, and Stone and Webster. The performance requirements for the PCAST machine are to form and sustain a burning plasma for three helium accumulation times. The philosophy adopted for this design was to achieve the required performance at lower cost by decreasing the major radius to five meters, increasing the toroidal field to 7 tesla, and using stronger shaping. The major device parameters are given. 4 refs., 4 figs.,more » 1 tab.« less
Fusion Science and Technology, 2013
Abstract The ITER project baseline now includes two sets of in-vessel coils, one to mitigate the ... more Abstract The ITER project baseline now includes two sets of in-vessel coils, one to mitigate the effects of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location presents special challenges in terms of nuclear radiation and temperature, and requires the use of mineral-insulated conductors. An update to the preliminary design based on this conductor technology is presented for both coil designs. Results from an on-going R&D program consisting of conductor development, welding and brazing process development, electrical testing and mechanical testing in order to demonstrate manufacturability of this style of conductor are presented. Plans for two prototype coils, one of each type, are presented.
Proceedings of the 2003 Bipolar/BiCMOS Circuits and Technology Meeting (IEEE Cat. No.03CH37440)
Fusion Engineering and Design
IEEE Transactions on Plasma Science
a successful conceptual design was completed to develop in-vessel control coils (Non-Axisymmetric... more a successful conceptual design was completed to develop in-vessel control coils (Non-Axisymmetric Control Coils or NCC). The NCC coils are a series of saddle coils that are intended to satisfy a number of physics criterion including magnetic breaking, error field control, fast Resistive Wall Mode (RWM) control and ELM stabilization. Customized Mineral Insulated Cable (MIC) was selected for the conductor material. The MIC was made from oxygen-free copper conductor, high purity Magnesium Oxide powder insulation and 304 stainless steel sleeve cover. Sample MIC was purchased from an outside supplier and various tests were conducted for design and performance verification including high voltage testing, which necessitated the development of special test terminations. A concept design was also developed for terminating the MIC ends using non-conductive and vacuum sealed technique. Two alternative designs were proposed for joints inside the vacuum vessel. The NCC Coils are designed to be mounted in-front of the primary passive plates and underneath the Plasma Facing Components (PFC) tiles. The passive plates will be modified to accommodate the coils. A new PFC tiles design concept was developed using High-Z materials. New designs for the penetrations through the vacuum vessel wall were developed. One port per coil was planned. A new power patch panel will be required to provide the ability for various combinations of connections between the NCC, the existing RWM Coils and the existing Switching Power Amplifier (SPA) power suppliers.
This is a preprint of a paper intended for presentation at a scientific meeting. Because of the p... more This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its content and since changes of substance or detail may have to be made before publication, the preprint is made available with the understanding that it not be cited in the literature or in any way be reproduced in its present form. The views expressed and the statements made remain the responsibility of the named author(s); the views do not necessarily reflect those of the government of the designating Member States(s) or of the designating organization(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material
Analyses of the plasma heating and current drive configuration and aspects of the tokamak configu... more Analyses of the plasma heating and current drive configuration and aspects of the tokamak configuration for K-DEMO are reported. A combination of heating and current drive technologies, i.e., lower hybrid waves (LH), neutral beams (NB), electron cyclotron waves (EC), and ion cyclotron conventional fast waves (IC) is considered for satisfying multiple requirements, including steady state current drive. Parameter scan calculations for each of the four technologies have quantified the dependence of key performance metrics, such as driven current per watt and radial profile, on design variables. Tokamak configuration studies have investigated requirements on the in-vessel systems for maintainability, access for piping services, and structural support against magnetic forces. The blanket-shield system is segmented into a small number (48) of large modules, supported in part by a segmented semi-permanent inboard shell. Structural analysis has addressed stresses due to eddy currents result...
Fusion Science and Technology
Fusion Engineering and Design
MIT, PSFC is in the process of designing and installing a cryopump in the Alcator C-MOD vessel. S... more MIT, PSFC is in the process of designing and installing a cryopump in the Alcator C-MOD vessel. Similar to that used at D-III-D [2], the cryopump will be used for density and impurity control during h-mode, ITB, and lower hybrid current drive operation on C-MOD. Performance goals are discussed in a companion paper at this conference on testing and installation. The mechanical design and analysis of the pump is presented in this paper. Significant loading includes differential contraction, pressurization due to up to to air fault in the vessel, disruption eddy currents, and halo currents in the tile shelf and supporting structures. The cold component support system is described and simulations of cooldown and pressurization are presented. Eddy currents are computed from a transient electromagnetic ANSYS simulation. This provides toroidal currents which in turn are input to a simulation of the pulse in which the cryopump disruption eddy currents are assumed to occur at any time in the...
2015 IEEE 26th Symposium on Fusion Engineering (SOFE), 2015
The Wendelstein 7-X (W7-X) stellarator is in final stages of commissioning, and will begin operat... more The Wendelstein 7-X (W7-X) stellarator is in final stages of commissioning, and will begin operation in the last half of 2015. In this first phase the machine will operate with a limiter, and will be restricted to low power and short pulse. But in 2019, plans are for an actively cooled divertor to be installed, and the machine will operate in steady-state at full power. Recently, plasma simulations have indicated that, in this final operational phase, a bootstrap current may evolve in certain scenarios. This will cause the sensitive ends of the divertor target to be overloaded beyond their qualified limit. A high heat flux scraper element (HHF-SE) has been proposed in order to take up some of the convective flux and reduce the load on the divertor. In order to examine whether the HHF-SE will be able to effectively reduce the plasma flux in the divertor region of concern, and to determine how the pumping effectiveness will be affected by such a component, it is planned to include a t...
Fusion Technology, 1996
A new type of magnet design is proposed, where the conductor is composed of conducting pebbles in... more A new type of magnet design is proposed, where the conductor is composed of conducting pebbles in a low-melting temperature conducting matrix. The magnet should have high radiation tolerance. At the end-of-life of the conductor, the pebbles can be circulated out of the magnet after the conducting matrix is molten. Application of this approach to the centerpost in the Low Aspect Ratio Device is discussed. 6 refs., 1 fig.
IEEE/NPSS Symposium on Fusion Engineering, Sep 30, 1991
Mandatory standards for magnet structural design have been proposed. They pertain to all phases o... more Mandatory standards for magnet structural design have been proposed. They pertain to all phases of magnet technology from conceptual design through lifetime performance, and are intended for all magnet types. The basis for development of the standards is presented, together with extracted sections of the structural guidelines document that preceded it, with a few examples to support conservatism in the design of magnet systems.<<ETX>>
Fusion Science and Technology, 2015
The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the co... more The proposed Fusion Nuclear Science Facility (FNSF) represents the first facility to enter the complex fusion nuclear regime, and its technical mission and attributes are being developed. The FNSF represents one part of the fusion energy development pathway to the first commercial power plant with other major components being the pre-FNSF research and development, research in parallel with the FNSF, pre-DEMO research and development, and the demonstration power plant (DEMO). The Fusion Energy Systems Studies group is developing the technical basis for the FNSF in order to provide a better understanding of the demands on the fusion plasma and fusion nuclear science programs. *higher β N operation can allow higher neutron wall load **1 year has been allocated at the end of each phase in addition to the specified maintenance time
Fusion Technology, 1996
The President`s Committee of Advisors on Science and Technology (PCAST) has suggested that a devi... more The President`s Committee of Advisors on Science and Technology (PCAST) has suggested that a device with a mission of ignition and moderate burn time could address the physics of burning plasmas at a lesser cost than ITER with its more comprehensive physics and technology mission. The Department of Energy commissioned a study to explore this PCAST suggestion. This paper describes the results of the engineering portion of the study of this `PCAST Machine;` physics is covered in a companion paper authored by G.H. Neilson, et al; and the costs are covered in a companion paper by R.T. Simmons, et al. Both are published in the proceedings of this conference. The study was undertaken by a team under the direction of Bruce Montgomery that included representatives from MIT, PPPL, ORNL, LLNL, GA, Northrup-Grumman, and Stone and Webster. The performance requirements for the PCAST machine are to form and sustain a burning plasma for three helium accumulation times. The philosophy adopted for this design was to achieve the required performance at lower cost by decreasing the major radius to five meters, increasing the toroidal field to 7 tesla, and using stronger shaping. The major device parameters are given. 4 refs., 4 figs.,more » 1 tab.« less
Fusion Science and Technology, 2013
Abstract The ITER project baseline now includes two sets of in-vessel coils, one to mitigate the ... more Abstract The ITER project baseline now includes two sets of in-vessel coils, one to mitigate the effects of Edge Localized Modes (ELMs) and another to provide vertical stabilization (VS). The in-vessel location presents special challenges in terms of nuclear radiation and temperature, and requires the use of mineral-insulated conductors. An update to the preliminary design based on this conductor technology is presented for both coil designs. Results from an on-going R&D program consisting of conductor development, welding and brazing process development, electrical testing and mechanical testing in order to demonstrate manufacturability of this style of conductor are presented. Plans for two prototype coils, one of each type, are presented.
Proceedings of the 2003 Bipolar/BiCMOS Circuits and Technology Meeting (IEEE Cat. No.03CH37440)
Fusion Engineering and Design
IEEE Transactions on Plasma Science
a successful conceptual design was completed to develop in-vessel control coils (Non-Axisymmetric... more a successful conceptual design was completed to develop in-vessel control coils (Non-Axisymmetric Control Coils or NCC). The NCC coils are a series of saddle coils that are intended to satisfy a number of physics criterion including magnetic breaking, error field control, fast Resistive Wall Mode (RWM) control and ELM stabilization. Customized Mineral Insulated Cable (MIC) was selected for the conductor material. The MIC was made from oxygen-free copper conductor, high purity Magnesium Oxide powder insulation and 304 stainless steel sleeve cover. Sample MIC was purchased from an outside supplier and various tests were conducted for design and performance verification including high voltage testing, which necessitated the development of special test terminations. A concept design was also developed for terminating the MIC ends using non-conductive and vacuum sealed technique. Two alternative designs were proposed for joints inside the vacuum vessel. The NCC Coils are designed to be mounted in-front of the primary passive plates and underneath the Plasma Facing Components (PFC) tiles. The passive plates will be modified to accommodate the coils. A new PFC tiles design concept was developed using High-Z materials. New designs for the penetrations through the vacuum vessel wall were developed. One port per coil was planned. A new power patch panel will be required to provide the ability for various combinations of connections between the NCC, the existing RWM Coils and the existing Switching Power Amplifier (SPA) power suppliers.
This is a preprint of a paper intended for presentation at a scientific meeting. Because of the p... more This is a preprint of a paper intended for presentation at a scientific meeting. Because of the provisional nature of its content and since changes of substance or detail may have to be made before publication, the preprint is made available with the understanding that it not be cited in the literature or in any way be reproduced in its present form. The views expressed and the statements made remain the responsibility of the named author(s); the views do not necessarily reflect those of the government of the designating Member States(s) or of the designating organization(s). In particular, neither the IAEA nor any other organization or body sponsoring this meeting can be held responsible for any material
Analyses of the plasma heating and current drive configuration and aspects of the tokamak configu... more Analyses of the plasma heating and current drive configuration and aspects of the tokamak configuration for K-DEMO are reported. A combination of heating and current drive technologies, i.e., lower hybrid waves (LH), neutral beams (NB), electron cyclotron waves (EC), and ion cyclotron conventional fast waves (IC) is considered for satisfying multiple requirements, including steady state current drive. Parameter scan calculations for each of the four technologies have quantified the dependence of key performance metrics, such as driven current per watt and radial profile, on design variables. Tokamak configuration studies have investigated requirements on the in-vessel systems for maintainability, access for piping services, and structural support against magnetic forces. The blanket-shield system is segmented into a small number (48) of large modules, supported in part by a segmented semi-permanent inboard shell. Structural analysis has addressed stresses due to eddy currents result...
Fusion Science and Technology
Fusion Engineering and Design