The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magneti... more The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and
A design study was undertaken to develop a ''first cut'' integrated mechanical de... more A design study was undertaken to develop a ''first cut'' integrated mechanical design concept of the final focusing region for a conceptual IFE power plant that considers the major issues which must be addressed in an integrated driver and chamber system. The conceptual design in this study requires a total of 120 beamlines located in two conical arrays attached on
The Energy Citations Database (ECD) provides access to historical and current research (1948 to t... more The Energy Citations Database (ECD) provides access to historical and current research (1948 to the present) from the Department of Energy (DOE) and predecessor agencies.
The National Compact Stellarator Experiment (NCSX) is the first of a new class of stellarators kn... more The National Compact Stellarator Experiment (NCSX) is the first of a new class of stellarators known as "compact stellarators". Stellarators are characterized by three dimensional magnetic fields and plasma shapes and are the best-developed class of magnetic fusion devices after the tokamak. Stellarators are attractive because they solve critical problems of magnetic fusion energy: steady state operation without current drive
The NSTX device plasma facing components [PFC] consist of inboard divertors, outboard divertors, ... more The NSTX device plasma facing components [PFC] consist of inboard divertors, outboard divertors, primary passive plates, secondary passive plates, a center stack casing (CSC), and the heating/cooling fluid distribution system. The PFC surfaces are protected by 3584 individually mounted carbon tiles. Surfaces exposed to high heat flux and/or high loads utilize composite C-C graphite and the remainder utilize less costly ATJ graphite. A variety of diagnostics are incorporated into the PFCs including thermocouples, Langmuir probes, Mirnov coils, and Rogowski coils. The NSTX device is designed to be operated in a pulse mode of five seconds on followed by five minutes off. Its PFC components are also required to be baked out to 350 °C. During operation the PFC tiles are permitted to ramp up thermally and then cool sufficiently between shots to prevent ratcheting during subsequent shots. The CSC tiles are required to be thermally isolated from the CSC so that the primary heat loss is radiation to the other PFC components. The other PFCs are thermally coupled to water cooled plates by conductive gaskets. Special mounts are required which permit thermal expansion and can withstand disruption loads while maintaining thermal contact. The tiles and mounts for the CSC are required to fall within a total radial space allotment of only 14 mm. A unique design for mounting graphite tiles to the CSC was developed which utilizes drift (shear) pins and Inconel brackets. Installation is accomplished via hidden fasteners accessed through very small holes in the tile faces. Analyses of the CSC mounting structure were performed and pull tests were performed on assemblies which simulated the attachment geometry in an attempt to determine the ultimate strength of the configuration and the mechanism of failure
ABSTRACT The Tokamak Physics Experiment (TPX) is a superconducting tokamak utilizing both Nb3Sn a... more ABSTRACT The Tokamak Physics Experiment (TPX) is a superconducting tokamak utilizing both Nb3Sn and NbTi superconducting magnets and will feature a low-activation titanium alloy vacuum vessel and carbon-carbon composite divertors. Due to the unique nature of the component designs, materials, and environment, the TPX project felt it necessary to develop a design criteria (code) which will specifically address the structural and cryogenic design aspects of such a device. The developed code is intended to serve all components of the device; namely, the TF and PF magnets, vacuum vessel, first wall and divertor, cryostat, diagnostics, heating devices, shielding, and all associated structural elements. The structural portion is based largely on that developed for the Burning Plasma Experiment (BPX), which was modeled after the CIT Vacuum Vessel Structural Design Criteria and ASME Boiler and Pressure Vessel (B&PV) Code. The cryogenic criteria is largely modeled after that proposed in the ITER CDA. This paper summarizes the TPX Criteria document
Proceedings of 16th International Symposium on Fusion Engineering, 1995
ABSTRACT PBX-M upgrades have been in progress for enhanced capability and reliability. These incl... more ABSTRACT PBX-M upgrades have been in progress for enhanced capability and reliability. These include upgrades to the vessel O-ring system for impurity control, the Ion Bernstein Wave and Lower Hybrid Current Drive Systems for plasma profile control, a double viewing upper and lower CHERS based Poloidal Rotation Diagnostic, and a new digital plasma control system
... Parameter Value Major radius, R,, 0.9 m Minor radius 0.3 - 0.4 m Bmax 1-1.2 T Plasma current,... more ... Parameter Value Major radius, R,, 0.9 m Minor radius 0.3 - 0.4 m Bmax 1-1.2 T Plasma current, lp up to 50 kA Auxiliary drive power, Pa, 0.9 MW @ 28 GHz 1 MW @ 56 GHz 2 MW @ 6-20 MHz i.5 MW @ 40-80 MHz 11. MODULAR COILS ... 71 mm x 142 mm 300 kA 380 kA 309 ...
The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magneti... more The Fusion Ignition Research Experiment (FIRE) is being designed as a next step in the US magnetic fusion program. The FIRE tokamak has a major radius of 2 m, a minor radius of 0.525 m, and liquid nitrogen cooled copper coils. The aim is to produce a pulse length of 20 s with a plasma current of 6.6 MA and
A design study was undertaken to develop a ''first cut'' integrated mechanical de... more A design study was undertaken to develop a ''first cut'' integrated mechanical design concept of the final focusing region for a conceptual IFE power plant that considers the major issues which must be addressed in an integrated driver and chamber system. The conceptual design in this study requires a total of 120 beamlines located in two conical arrays attached on
The Energy Citations Database (ECD) provides access to historical and current research (1948 to t... more The Energy Citations Database (ECD) provides access to historical and current research (1948 to the present) from the Department of Energy (DOE) and predecessor agencies.
The National Compact Stellarator Experiment (NCSX) is the first of a new class of stellarators kn... more The National Compact Stellarator Experiment (NCSX) is the first of a new class of stellarators known as "compact stellarators". Stellarators are characterized by three dimensional magnetic fields and plasma shapes and are the best-developed class of magnetic fusion devices after the tokamak. Stellarators are attractive because they solve critical problems of magnetic fusion energy: steady state operation without current drive
The NSTX device plasma facing components [PFC] consist of inboard divertors, outboard divertors, ... more The NSTX device plasma facing components [PFC] consist of inboard divertors, outboard divertors, primary passive plates, secondary passive plates, a center stack casing (CSC), and the heating/cooling fluid distribution system. The PFC surfaces are protected by 3584 individually mounted carbon tiles. Surfaces exposed to high heat flux and/or high loads utilize composite C-C graphite and the remainder utilize less costly ATJ graphite. A variety of diagnostics are incorporated into the PFCs including thermocouples, Langmuir probes, Mirnov coils, and Rogowski coils. The NSTX device is designed to be operated in a pulse mode of five seconds on followed by five minutes off. Its PFC components are also required to be baked out to 350 °C. During operation the PFC tiles are permitted to ramp up thermally and then cool sufficiently between shots to prevent ratcheting during subsequent shots. The CSC tiles are required to be thermally isolated from the CSC so that the primary heat loss is radiation to the other PFC components. The other PFCs are thermally coupled to water cooled plates by conductive gaskets. Special mounts are required which permit thermal expansion and can withstand disruption loads while maintaining thermal contact. The tiles and mounts for the CSC are required to fall within a total radial space allotment of only 14 mm. A unique design for mounting graphite tiles to the CSC was developed which utilizes drift (shear) pins and Inconel brackets. Installation is accomplished via hidden fasteners accessed through very small holes in the tile faces. Analyses of the CSC mounting structure were performed and pull tests were performed on assemblies which simulated the attachment geometry in an attempt to determine the ultimate strength of the configuration and the mechanism of failure
ABSTRACT The Tokamak Physics Experiment (TPX) is a superconducting tokamak utilizing both Nb3Sn a... more ABSTRACT The Tokamak Physics Experiment (TPX) is a superconducting tokamak utilizing both Nb3Sn and NbTi superconducting magnets and will feature a low-activation titanium alloy vacuum vessel and carbon-carbon composite divertors. Due to the unique nature of the component designs, materials, and environment, the TPX project felt it necessary to develop a design criteria (code) which will specifically address the structural and cryogenic design aspects of such a device. The developed code is intended to serve all components of the device; namely, the TF and PF magnets, vacuum vessel, first wall and divertor, cryostat, diagnostics, heating devices, shielding, and all associated structural elements. The structural portion is based largely on that developed for the Burning Plasma Experiment (BPX), which was modeled after the CIT Vacuum Vessel Structural Design Criteria and ASME Boiler and Pressure Vessel (B&PV) Code. The cryogenic criteria is largely modeled after that proposed in the ITER CDA. This paper summarizes the TPX Criteria document
Proceedings of 16th International Symposium on Fusion Engineering, 1995
ABSTRACT PBX-M upgrades have been in progress for enhanced capability and reliability. These incl... more ABSTRACT PBX-M upgrades have been in progress for enhanced capability and reliability. These include upgrades to the vessel O-ring system for impurity control, the Ion Bernstein Wave and Lower Hybrid Current Drive Systems for plasma profile control, a double viewing upper and lower CHERS based Poloidal Rotation Diagnostic, and a new digital plasma control system
... Parameter Value Major radius, R,, 0.9 m Minor radius 0.3 - 0.4 m Bmax 1-1.2 T Plasma current,... more ... Parameter Value Major radius, R,, 0.9 m Minor radius 0.3 - 0.4 m Bmax 1-1.2 T Plasma current, lp up to 50 kA Auxiliary drive power, Pa, 0.9 MW @ 28 GHz 1 MW @ 56 GHz 2 MW @ 6-20 MHz i.5 MW @ 40-80 MHz 11. MODULAR COILS ... 71 mm x 142 mm 300 kA 380 kA 309 ...
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