Dedicated to carrying out innovative research to advance fusion energy toward a practical, widely implemented, and environmentally friendly energy source. In the near-term, the R
Fully non-inductive plasma current start-up without the central solenoid in ECW plasma was used o... more Fully non-inductive plasma current start-up without the central solenoid in ECW plasma was used on EXL-50 Spherical Torus with a weak external vertical field (Bv). Generally, the number of electrons leaving to the vessel wall by the gradient Bt is larger than ions, and the positive potential was built up in plasma. The relationship between floating potential and the plasma current was studied using the Langmuir probes near the boundary. The results show that the floating potential is positive (about 200V) and has a strong correlation with plasma current. In open magnetic field, the plasma current is driven by the high energy electrons in preferential confinement, the plasma current and potential approximately positively correlated with total electron density. After forming the closed flux surface, the plasma current consists mainly of the ECW driven current, and potential is negatively correlated with plasma current. By actively adjusting the Bv, it demonstrated that the positive voltage is approximately inversely correlated with the Bv and plasma current (Ip). Considering that the plasma temperature near the boundary is quite low (~eV), the positive voltage near the boundary caused by the high-energy electron loss. Therefore, the measurements of the boundary potential are important for the study of high-energy electron confinement performance, noninductive plasma current start-up and current driven.
The use of a fusion component testing facility to study and establish during the ITER era the rem... more The use of a fusion component testing facility to study and establish during the ITER era the remaining scientific and technical knowledge needed by fusion Demo is considered and described in this paper. This use aims to test components in an integrated fusion nuclear environment to discover and understand the underpinning physical properties, and to develop improved components for further testing, in a time-efficient manner. It requires a design with extensive modularization and remote handling of activated components, and flexible hot-cell laboratories. It further requires reliable plasma conditions to avoid disruptions and minimize impact, and designs to reduce the divertor heat flux. As the plasma duration is extended through the ITER level (~10 s) and beyond, physical properties with increasing time constants would become accessible for testing and R&D. The longest time constants of these are expected to be many days (~10 – 10 s). Progressive stages of research operation are en...
The goal of stability and steady-state control research is to generate physics understanding of h... more The goal of stability and steady-state control research is to generate physics understanding of high beta plasma stability that will produce reliable high performance, continuous plasma operation with both negligible stored energy and neutron output fluctuation, and determination of optimal plasma operating conditions and control systems allowing confident extrapolation to future tokamak and ST burning plasmas (spanning from upgrades of present devices to DEMO-level plasma conditions).
Research description Mastering the science of plasma material interactions (PMI) and the technolo... more Research description Mastering the science of plasma material interactions (PMI) and the technology of plasma facing components (PFCs) will be key to the successful exploitation of fusion energy. PFCs must be capable of exhausting the power leaving the core plasma (heat fluxes of ~5-10 MW/m 2), while avoiding excessive net deposition or erosion of the material surface in the presence of high ion fluxes (Γ~10 24 m-2 s-1). The continual gross erosion and redeposition will lead to surface layers that have been strongly modified by the plasma, which will have properties that are, at present, unknown. Further, control over the tritium inventory must be ensured, and performance maintained in the presence of intense neutron irradiation (on the order of ~50 dpa (displacements per atom)). Finally, this must be accomplished in a way compatible with good core plasma performance. PFC performance depends on both the material properties and on the characteristics of the impinging plasma, and a combination which satisfies these requirements has yet to be identified. A facility dedicated to improving the understanding of PMI physics, exploring potential PFC materials, testing material limits, and developing the associated divertor and first wall technology will reduce the risk and improve the designs of ITER, FNSF, DEMO and beyond. Such a facility must be capable of producing a plasma with a wide range of parameters near the PFC surface. Taking the ITER divertor scenario as an example, conditions vary from a 'detached', cold (T e ~ 1 eV) very dense (n e > 2 x 10 21 m-3) plasma at the strike point to a hotter (5 < T e < 50 eV) 'attached' plasma with reduced density (10 21 > n e > 10 19 m-3) a short distance into the scrape-off layer (SOL). The detached strike point region is expected to be one of net deposition (because physical sputtering is minimal). While the lack of net erosion at these plasma parameters may be attractive (although the deposition would lead to surface morphology changes and possibly problems with flaking or poor heat transfer if the new layers are thick enough), tokamak operation experience has excluded a fully detached divertor as incompatible with sufficient core confinement, and so a hotter attached region is unavoidable. This region is expected to be a net erosion zone, which could limit the PFC lifetime. To evaluate a potential divertor material such as tungsten the erosion/redepostion mechanisms must be quantified and understood, and effective redeposition (of order 99% for high-Z materials) demonstrated. Much of the needed PMI studies and PFC development could be performed in a simplified geometry (i.e., in linear plasma devices), provided that the needed plasma parameters can be reached. These devices offer a much reduced cost compared to tokamak operation, with better diagnostic access and dedicated experimental time for PMI/PFC studies. The prospects of operating linear plasma devices in near steady-state to perform tests at reactor-relevant ion fluences are also much better than for tokamaks (current pulsed tokamaks accumulate a fluence of ~10 25 m-2 /yr only, which is about 5 orders of magnitude below what is needed). Such a device would allow rapid evaluation and the development of combinations of PFC designs and plasma conditions that satisfy the requirements listed above. Development of PFCs beyond the divertor components would also be enabled, such as studying particle-antenna interactions and improving RF launcher design. Testing in tokamaks will of course be necessary as well, especially to
The Large Aspect Ratio Tokanak Study (LARTS) investigated the potential for producing a viable lo... more The Large Aspect Ratio Tokanak Study (LARTS) investigated the potential for producing a viable long burn tokanak reactor through enhsnced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering; and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohnic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured. 'Research sponsored by the Office of Fusion Energy, U.S. Department of Energy under contract W-7405-eng-26 with the Union Carbide Corporation. •v Kcipanct at th.i »;Tid«. th» publ'thtr O' rtc.pitnt •c'tnowltdftl tno l.S. Gwv««m«nri n(.u to r«»in • tontxclufivt, rovJftvf'M licatM in and TO M*y copyifht
Los Alames National Labomtocy, an affirmativeaction/equalopportunityem_k:_y_', is ogerated by the... more Los Alames National Labomtocy, an affirmativeaction/equalopportunityem_k:_y_', is ogerated by the Universityof Californiafor the U.S. Department of Energy under contractW-7405-ENG-38. By acceptar_ of thisarticle,the publisherrecognizesthat the U.S. Government retainsa nonexclusive,royalty-free license to publish or reproducethe publishedformof thiscontribution, or to allow other_to do Io, forU,S. Governmentpurposes.The LosAlamos NationalLaboratory requeststhat the Ix_C'_r identifyth_ articleas work performedunderthe auN:k:es ofthe U.S. Departmentof Energy. FormNo.836 R5 ST2629 _0R_
This report «as prepared a* an accoant of wort sponsored by aa agency of the United States Govcra... more This report «as prepared a* an accoant of wort sponsored by aa agency of the United States Govcrameat. Neither the Uaitcd Sutes Gotciamcat aor «ay agency thereof, «or aay of their ORNL/FEDC-83/3 employees, makes any warranty, express or implied, or escames aay legal lability or respoasi-r,i s t Category UC-20 C, d baity for the acoaracy, completeness, or aaefalfss of aay mfonnation, appararat, prodact, or process ajsdoscd, or reprtacnü that its ase woaid apt infringe pmaicty owaed rights. Refereace herein to aay specific commercial prodact, process, or «enrice by trade aame, trademark, maaafactarer, or otherwise does not necessarily cuastitate or imply its endorsement, recom-mradatioB, or favoring by the United States GoTenuneat or any agency thereof. The views and opinions of aathors expressed bereia do not n-jcessariry state or reflect those of the United States Government or aay agency thereof.
IIII a s 1113 tl|/£rlM(5i2E y CADR^W!* Research sponsored by the Office of Fusion Energy, U.S. De... more IIII a s 1113 tl|/£rlM(5i2E y CADR^W!* Research sponsored by the Office of Fusion Energy, U.S. Department of Energy, under ,*-«p| contract DE-AC05-840R21400 with Martin Hf^flfK Marietta Energy Systems, Incorporated. wflofj'^* OtSTRIEUTiON OF THIS DOCUMENT (6 UNLIIVJJTEH OF 2. 3. EXAMPi£ OF MODULES 4. ftEUWfjAPf mOLJ<; Ol TfdER 2t FUSION ENGINEERING DESIGN CENTER
The ISX-B tokamak h a s a p o l o i d a l c o i l system designed t o produce c i r c u l a r , e... more The ISX-B tokamak h a s a p o l o i d a l c o i l system designed t o produce c i r c u l a r , e l l i p t i c a l , and D-shaped plasmas.
Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction... more Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal non-rotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants.
This report was prepared as an account of work sponsored by an agency of the United States Govern... more This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility for any third party's use or the results of such use of any information, apparatus, product or process disclosed in this report, nor represents that its use by such third party would not infringe privately owned rights.
The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Labor... more The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, fhe first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.
A major purpose of the Techni cal Information Center is to provide the broadest dissemination pos... more A major purpose of the Techni cal Information Center is to provide the broadest dissemination possi ble of information contained in DOE's Research and Development Reports to business, industry, the academic community, and federal, state and local governments. Although a small portion of this report is not reproducible, it is being made available to expedite the availability of information on the research discussed herein.
This report was prepared as an account of work sponsored by an agency of the United StatesGovernm... more This report was prepared as an account of work sponsored by an agency of the United StatesGovernment. Neitherthe UnitedStatesGovernment norany agency thereof, nor any of their employees, contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility for any third party's use or the results of such use of any information, apparatus, product or process disclosed in this report, nor represents that its use by such third party would not infringe privately owned rights.
NOTICE This document contains information of a preliminary nature and was praparad priri~arily fo... more NOTICE This document contains information of a preliminary nature and was praparad priri~arily for Internal use at the Oak Ridge National Laboratory. I t is subject t o revision or correction and therefore does not represent a final report.
This book wat of*pw«t »s an account of work sponsored by an agency ol ttie United Staiet G'jf nme... more This book wat of*pw«t »s an account of work sponsored by an agency ol ttie United Staiet G'jf nmem. Neither the United Stale* Government nor any agency thereof, nor anv of the'' employee}, make! any warranty, eipreu or Implied, or aiwmet any lagal liabiliiv or reipamiCil'ty (or Hie accuracy. comutetfrrmt, o< u«iuln«« o (tny «n!orm«foof apiwawi. ptotiuci. of P'QCHII disdased. or repreMnii thai its me vould noi infringe Ofivilely owned rightv Pclerenco herein to any tpeci'ic commercial uroduci, pern*, or vrvice by trade rum*, trademark, minute'.:!urer, or otherwise, does not neccudrlly to mil lute or imply its endorsement, recommendation, or favoring by itw United StaT« Goytfntn^l or a^Y SQBncy trwteo'. The views and opinion) of aulhorl expressed herein da noi i|y state or rellecl ihoia at the United Sutei Government or any ajency fr-eieot.
Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Con... more Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes n ≈ 8.8 × 10 13 n s −1 cm −2 ('wall loading' W L ≈ 2 MW m −2), over size-scale >10 5 cm 2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' > 0.3 MW yr −1 m −2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R 0 = 1.2 m, A = 1.5, elongation ∼3, I p ∼ 12 MA, B T ∼ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere
18,24 and for the International ITB Database Working Group and the responsible officers for the I... more 18,24 and for the International ITB Database Working Group and the responsible officers for the ITPA collaborative experiments on the 'hybrid' and 'steady-state' regimes:
Fully non-inductive plasma current start-up without the central solenoid in ECW plasma was used o... more Fully non-inductive plasma current start-up without the central solenoid in ECW plasma was used on EXL-50 Spherical Torus with a weak external vertical field (Bv). Generally, the number of electrons leaving to the vessel wall by the gradient Bt is larger than ions, and the positive potential was built up in plasma. The relationship between floating potential and the plasma current was studied using the Langmuir probes near the boundary. The results show that the floating potential is positive (about 200V) and has a strong correlation with plasma current. In open magnetic field, the plasma current is driven by the high energy electrons in preferential confinement, the plasma current and potential approximately positively correlated with total electron density. After forming the closed flux surface, the plasma current consists mainly of the ECW driven current, and potential is negatively correlated with plasma current. By actively adjusting the Bv, it demonstrated that the positive voltage is approximately inversely correlated with the Bv and plasma current (Ip). Considering that the plasma temperature near the boundary is quite low (~eV), the positive voltage near the boundary caused by the high-energy electron loss. Therefore, the measurements of the boundary potential are important for the study of high-energy electron confinement performance, noninductive plasma current start-up and current driven.
The use of a fusion component testing facility to study and establish during the ITER era the rem... more The use of a fusion component testing facility to study and establish during the ITER era the remaining scientific and technical knowledge needed by fusion Demo is considered and described in this paper. This use aims to test components in an integrated fusion nuclear environment to discover and understand the underpinning physical properties, and to develop improved components for further testing, in a time-efficient manner. It requires a design with extensive modularization and remote handling of activated components, and flexible hot-cell laboratories. It further requires reliable plasma conditions to avoid disruptions and minimize impact, and designs to reduce the divertor heat flux. As the plasma duration is extended through the ITER level (~10 s) and beyond, physical properties with increasing time constants would become accessible for testing and R&D. The longest time constants of these are expected to be many days (~10 – 10 s). Progressive stages of research operation are en...
The goal of stability and steady-state control research is to generate physics understanding of h... more The goal of stability and steady-state control research is to generate physics understanding of high beta plasma stability that will produce reliable high performance, continuous plasma operation with both negligible stored energy and neutron output fluctuation, and determination of optimal plasma operating conditions and control systems allowing confident extrapolation to future tokamak and ST burning plasmas (spanning from upgrades of present devices to DEMO-level plasma conditions).
Research description Mastering the science of plasma material interactions (PMI) and the technolo... more Research description Mastering the science of plasma material interactions (PMI) and the technology of plasma facing components (PFCs) will be key to the successful exploitation of fusion energy. PFCs must be capable of exhausting the power leaving the core plasma (heat fluxes of ~5-10 MW/m 2), while avoiding excessive net deposition or erosion of the material surface in the presence of high ion fluxes (Γ~10 24 m-2 s-1). The continual gross erosion and redeposition will lead to surface layers that have been strongly modified by the plasma, which will have properties that are, at present, unknown. Further, control over the tritium inventory must be ensured, and performance maintained in the presence of intense neutron irradiation (on the order of ~50 dpa (displacements per atom)). Finally, this must be accomplished in a way compatible with good core plasma performance. PFC performance depends on both the material properties and on the characteristics of the impinging plasma, and a combination which satisfies these requirements has yet to be identified. A facility dedicated to improving the understanding of PMI physics, exploring potential PFC materials, testing material limits, and developing the associated divertor and first wall technology will reduce the risk and improve the designs of ITER, FNSF, DEMO and beyond. Such a facility must be capable of producing a plasma with a wide range of parameters near the PFC surface. Taking the ITER divertor scenario as an example, conditions vary from a 'detached', cold (T e ~ 1 eV) very dense (n e > 2 x 10 21 m-3) plasma at the strike point to a hotter (5 < T e < 50 eV) 'attached' plasma with reduced density (10 21 > n e > 10 19 m-3) a short distance into the scrape-off layer (SOL). The detached strike point region is expected to be one of net deposition (because physical sputtering is minimal). While the lack of net erosion at these plasma parameters may be attractive (although the deposition would lead to surface morphology changes and possibly problems with flaking or poor heat transfer if the new layers are thick enough), tokamak operation experience has excluded a fully detached divertor as incompatible with sufficient core confinement, and so a hotter attached region is unavoidable. This region is expected to be a net erosion zone, which could limit the PFC lifetime. To evaluate a potential divertor material such as tungsten the erosion/redepostion mechanisms must be quantified and understood, and effective redeposition (of order 99% for high-Z materials) demonstrated. Much of the needed PMI studies and PFC development could be performed in a simplified geometry (i.e., in linear plasma devices), provided that the needed plasma parameters can be reached. These devices offer a much reduced cost compared to tokamak operation, with better diagnostic access and dedicated experimental time for PMI/PFC studies. The prospects of operating linear plasma devices in near steady-state to perform tests at reactor-relevant ion fluences are also much better than for tokamaks (current pulsed tokamaks accumulate a fluence of ~10 25 m-2 /yr only, which is about 5 orders of magnitude below what is needed). Such a device would allow rapid evaluation and the development of combinations of PFC designs and plasma conditions that satisfy the requirements listed above. Development of PFCs beyond the divertor components would also be enabled, such as studying particle-antenna interactions and improving RF launcher design. Testing in tokamaks will of course be necessary as well, especially to
The Large Aspect Ratio Tokanak Study (LARTS) investigated the potential for producing a viable lo... more The Large Aspect Ratio Tokanak Study (LARTS) investigated the potential for producing a viable long burn tokanak reactor through enhsnced volt-second capability of the ohmic heating transformer by employing high aspect ratio designs. The plasma physics, engineering; and economic implications of high aspect ratio tokamaks were accessed in the context of extended burn operation. Plasma startup and burn parameters were addressed using a one-dimensional transport code. The pulsed electrical power requirements for the poloidal field system, which have a major impact on reactor economics, were minimized by optimizing the field in the ohnic heating coil and the wave shape of the ohmic heating discharge. A high aspect ratio reference reactor was chosen and configured. 'Research sponsored by the Office of Fusion Energy, U.S. Department of Energy under contract W-7405-eng-26 with the Union Carbide Corporation. •v Kcipanct at th.i »;Tid«. th» publ'thtr O' rtc.pitnt •c'tnowltdftl tno l.S. Gwv««m«nri n(.u to r«»in • tontxclufivt, rovJftvf'M licatM in and TO M*y copyifht
Los Alames National Labomtocy, an affirmativeaction/equalopportunityem_k:_y_', is ogerated by the... more Los Alames National Labomtocy, an affirmativeaction/equalopportunityem_k:_y_', is ogerated by the Universityof Californiafor the U.S. Department of Energy under contractW-7405-ENG-38. By acceptar_ of thisarticle,the publisherrecognizesthat the U.S. Government retainsa nonexclusive,royalty-free license to publish or reproducethe publishedformof thiscontribution, or to allow other_to do Io, forU,S. Governmentpurposes.The LosAlamos NationalLaboratory requeststhat the Ix_C'_r identifyth_ articleas work performedunderthe auN:k:es ofthe U.S. Departmentof Energy. FormNo.836 R5 ST2629 _0R_
This report «as prepared a* an accoant of wort sponsored by aa agency of the United States Govcra... more This report «as prepared a* an accoant of wort sponsored by aa agency of the United States Govcrameat. Neither the Uaitcd Sutes Gotciamcat aor «ay agency thereof, «or aay of their ORNL/FEDC-83/3 employees, makes any warranty, express or implied, or escames aay legal lability or respoasi-r,i s t Category UC-20 C, d baity for the acoaracy, completeness, or aaefalfss of aay mfonnation, appararat, prodact, or process ajsdoscd, or reprtacnü that its ase woaid apt infringe pmaicty owaed rights. Refereace herein to aay specific commercial prodact, process, or «enrice by trade aame, trademark, maaafactarer, or otherwise does not necessarily cuastitate or imply its endorsement, recom-mradatioB, or favoring by the United States GoTenuneat or any agency thereof. The views and opinions of aathors expressed bereia do not n-jcessariry state or reflect those of the United States Government or aay agency thereof.
IIII a s 1113 tl|/£rlM(5i2E y CADR^W!* Research sponsored by the Office of Fusion Energy, U.S. De... more IIII a s 1113 tl|/£rlM(5i2E y CADR^W!* Research sponsored by the Office of Fusion Energy, U.S. Department of Energy, under ,*-«p| contract DE-AC05-840R21400 with Martin Hf^flfK Marietta Energy Systems, Incorporated. wflofj'^* OtSTRIEUTiON OF THIS DOCUMENT (6 UNLIIVJJTEH OF 2. 3. EXAMPi£ OF MODULES 4. ftEUWfjAPf mOLJ<; Ol TfdER 2t FUSION ENGINEERING DESIGN CENTER
The ISX-B tokamak h a s a p o l o i d a l c o i l system designed t o produce c i r c u l a r , e... more The ISX-B tokamak h a s a p o l o i d a l c o i l system designed t o produce c i r c u l a r , e l l i p t i c a l , and D-shaped plasmas.
Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction... more Ideal magnetohydrodynamic stability limits of shaped tokamak plasmas with high bootstrap fraction are systematically determined as a function of plasma aspect ratio. For plasmas with and without wall stabilization of external kink modes, the computed limits are well described by distinct and nearly invariant values of a normalized beta parameter utilizing the total magnetic field energy density inside the plasma. Stability limit data from the low aspect ratio National Spherical Torus Experiment is compared to these theoretical limits and indicates that ideal non-rotating plasma no-wall beta limits have been exceeded in regimes with sufficiently high cylindrical safety factor. These results could impact the choice of aspect ratio in future fusion power plants.
This report was prepared as an account of work sponsored by an agency of the United States Govern... more This report was prepared as an account of work sponsored by an agency of the United States Government. Neither the United States Government nor any agency thereof, nor any of their employees, contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility for any third party's use or the results of such use of any information, apparatus, product or process disclosed in this report, nor represents that its use by such third party would not infringe privately owned rights.
The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Labor... more The NSTX (National Spherical Torus Experiment) facility located at Princeton Plasma Physics Laboratory is the newest national fusion science experimental facility for the restructured US Fusion Energy Science Program. The NSTX project was approved in FY 97 as the first proof-of-principle national fusion facility dedicated to the spherical torus research. On Feb. 15, 1999, fhe first plasma was achieved 10 weeks ahead of schedule. The project was completed on budget and with an outstanding safety record. This paper gives an overview of the NSTX facility construction and the initial plasma operations.
A major purpose of the Techni cal Information Center is to provide the broadest dissemination pos... more A major purpose of the Techni cal Information Center is to provide the broadest dissemination possi ble of information contained in DOE's Research and Development Reports to business, industry, the academic community, and federal, state and local governments. Although a small portion of this report is not reproducible, it is being made available to expedite the availability of information on the research discussed herein.
This report was prepared as an account of work sponsored by an agency of the United StatesGovernm... more This report was prepared as an account of work sponsored by an agency of the United StatesGovernment. Neitherthe UnitedStatesGovernment norany agency thereof, nor any of their employees, contractors, subcontractors, or their employees, makes any warranty, express or implied, nor assumes any legal liability or responsibility for any third party's use or the results of such use of any information, apparatus, product or process disclosed in this report, nor represents that its use by such third party would not infringe privately owned rights.
NOTICE This document contains information of a preliminary nature and was praparad priri~arily fo... more NOTICE This document contains information of a preliminary nature and was praparad priri~arily for Internal use at the Oak Ridge National Laboratory. I t is subject t o revision or correction and therefore does not represent a final report.
This book wat of*pw«t »s an account of work sponsored by an agency ol ttie United Staiet G'jf nme... more This book wat of*pw«t »s an account of work sponsored by an agency ol ttie United Staiet G'jf nmem. Neither the United Stale* Government nor any agency thereof, nor anv of the'' employee}, make! any warranty, eipreu or Implied, or aiwmet any lagal liabiliiv or reipamiCil'ty (or Hie accuracy. comutetfrrmt, o< u«iuln«« o (tny «n!orm«foof apiwawi. ptotiuci. of P'QCHII disdased. or repreMnii thai its me vould noi infringe Ofivilely owned rightv Pclerenco herein to any tpeci'ic commercial uroduci, pern*, or vrvice by trade rum*, trademark, minute'.:!urer, or otherwise, does not neccudrlly to mil lute or imply its endorsement, recommendation, or favoring by itw United StaT« Goytfntn^l or a^Y SQBncy trwteo'. The views and opinion) of aulhorl expressed herein da noi i|y state or rellecl ihoia at the United Sutei Government or any ajency fr-eieot.
Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Con... more Recent experiments (Synakowski et al 2004 Nucl. Fusion 43 1648, Lloyd et al 2004 Plasma Phys. Control. Fusion 46 B477) on the Spherical Tokamak (or Spherical Torus, ST) (Peng 2000 Phys. Plasmas 7 1681) have discovered robust plasma conditions, easing shaping, stability limits, energy confinement, self-driven current and sustainment. This progress has encouraged an update of the plasma conditions and engineering of a Component Test Facility (CTF), (Cheng 1998 Fusion Eng. Des. 38 219) which is a very valuable step in the development of practical fusion energy. The testing conditions in a CTF are characterized by high fusion neutron fluxes n ≈ 8.8 × 10 13 n s −1 cm −2 ('wall loading' W L ≈ 2 MW m −2), over size-scale >10 5 cm 2 and depth-scale >50 cm, delivering >3 accumulated displacement per atom per year ('neutron fluence' > 0.3 MW yr −1 m −2) (Abdou et al 1999 Fusion Technol. 29 1). Such conditions are estimated to be achievable in a CTF with R 0 = 1.2 m, A = 1.5, elongation ∼3, I p ∼ 12 MA, B T ∼ 2.5 T, producing a driven fusion burn using 47 MW of combined neutral beam and RF heating power. A design concept that allows straight-line access via remote handling to all activated fusion core components is developed and presented. The ST CTF will test the lifetime of single-turn, copper alloy centre leg for the toroidal field coil without an induction solenoid and neutron shielding and require physics data on solenoid-free plasma current initiation, ramp-up to and sustainment at multiple megaampere
18,24 and for the International ITB Database Working Group and the responsible officers for the I... more 18,24 and for the International ITB Database Working Group and the responsible officers for the ITPA collaborative experiments on the 'hybrid' and 'steady-state' regimes:
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