Progress Toward Attractive Stellarators

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neilson, G H; Brown, T G; Gates, D A

The quasi-axisymmetric stellarator (QAS) concept offers a promising path to a more compact stellarator reactor, closer in linear dimensions to tokamak reactors than previous stellarator designs. Concept improvements are needed, however, to make it more maintainable and more compatible with high plant availability. Using the ARIES-CS design as a starting point, compact stellarator designs with improved maintenance characteristics have been developed. While the ARIES-CS features a through-the-port maintenance scheme, we have investigated configuration changes to enable a sector-maintenance approach, as envisioned for example in ARIES AT. Three approaches are reported. The first is to make tradeoffs within the QAS designmore » space, giving greater emphasis to maintainability criteria. The second approach is to improve the optimization tools to more accurately and efficiently target the physics properties of importance. The third is to employ a hybrid coil topology, so that the plasma shaping functions of the main coils are shared more optimally, either with passive conductors made of high-temperature superconductor or with local compensation coils, allowing the main coils to become simpler. Optimization tools are being improved to test these approaches.« less

ADX: a high field, high power density, advanced divertor and RF tokamak

NASA Astrophysics Data System (ADS)

LaBombard, B.; Marmar, E.; Irby, J.; Terry, J. L.; Vieira, R.; Wallace, G.; Whyte, D. G.; Wolfe, S.; Wukitch, S.; Baek, S.; Beck, W.; Bonoli, P.; Brunner, D.; Doody, J.; Ellis, R.; Ernst, D.; Fiore, C.; Freidberg, J. P.; Golfinopoulos, T.; Granetz, R.; Greenwald, M.; Hartwig, Z. S.; Hubbard, A.; Hughes, J. W.; Hutchinson, I. H.; Kessel, C.; Kotschenreuther, M.; Leccacorvi, R.; Lin, Y.; Lipschultz, B.; Mahajan, S.; Minervini, J.; Mumgaard, R.; Nygren, R.; Parker, R.; Poli, F.; Porkolab, M.; Reinke, M. L.; Rice, J.; Rognlien, T.; Rowan, W.; Shiraiwa, S.; Terry, D.; Theiler, C.; Titus, P.; Umansky, M.; Valanju, P.; Walk, J.; White, A.; Wilson, J. R.; Wright, G.; Zweben, S. J.

2015-05-01

The MIT Plasma Science and Fusion Center and collaborators are proposing a high-performance Advanced Divertor and RF tokamak eXperiment (ADX)—a tokamak specifically designed to address critical gaps in the world fusion research programme on the pathway to next-step devices: fusion nuclear science facility (FNSF), fusion pilot plant (FPP) and/or demonstration power plant (DEMO). This high-field (⩾6.5 T, 1.5 MA), high power density facility (P/S ˜ 1.5 MW m-2) will test innovative divertor ideas, including an ‘X-point target divertor’ concept, at the required performance parameters—reactor-level boundary plasma pressures, magnetic field strengths and parallel heat flux densities entering into the divertor region—while simultaneously producing high-performance core plasma conditions that are prototypical of a reactor: equilibrated and strongly coupled electrons and ions, regimes with low or no torque, and no fuelling from external heating and current drive systems. Equally important, the experimental platform will test innovative concepts for lower hybrid current drive and ion cyclotron range of frequency actuators with the unprecedented ability to deploy launch structures both on the low-magnetic-field side and the high-magnetic-field side—the latter being a location where energetic plasma-material interactions can be controlled and favourable RF wave physics leads to efficient current drive, current profile control, heating and flow drive. This triple combination—advanced divertors, advanced RF actuators, reactor-prototypical core plasma conditions—will enable ADX to explore enhanced core confinement physics, such as made possible by reversed central shear, using only the types of external drive systems that are considered viable for a fusion power plant. Such an integrated demonstration of high-performance core-divertor operation with steady-state sustainment would pave the way towards an attractive pilot plant, as envisioned in the ARC concept (affordable, robust, compact) (Sorbom et al 2015 Fusion Eng. Des. submitted (arXiv:1409.3540)) that makes use of high-temperature superconductor technology—a high-field (9.25 T) tokamak the size of the Joint European Torus that produces 270 MW of net electricity.

Alternative approaches to plasma confinement

NASA Technical Reports Server (NTRS)

Roth, J. R.

1977-01-01

The potential applications of fusion reactors, the desirable properties of reactors intended for various applications, and the limitations of the Tokamak concept are discussed. The principles and characteristics of 20 distinct alternative confinement concepts are described, each of which may be an alternative to the Tokamak. The devices are classed as Tokamak-like, stellarator-like, mirror machines, bumpy tori, electrostatically assisted, migma concept, and wall-confined plasma.

ARC: A compact, high-field, disassemblable fusion nuclear science facility and demonstration power plant

NASA Astrophysics Data System (ADS)

Sorbom, Brandon; Ball, Justin; Palmer, Timothy; Mangiarotti, Franco; Sierchio, Jennifer; Bonoli, Paul; Kasten, Cale; Sutherland, Derek; Barnard, Harold; Haakonsen, Christian; Goh, Jon; Sung, Choongki; Whyte, Dennis

2014-10-01

The Affordable, Robust, Compact (ARC) reactor conceptual design aims to reduce the size, cost, and complexity of a combined Fusion Nuclear Science Facility (FNSF) and demonstration fusion pilot power plant. ARC is a 270 MWe tokamak reactor with a major radius of 3.3 m, a minor radius of 1.1 m, and an on-axis magnetic field of 9.2 T. ARC has Rare Earth Barium Copper Oxide (REBCO) superconducting toroidal field coils with joints to allow disassembly, allowing for removal and replacement of the vacuum vessel as a single component. Inboard-launched current drive of 25 MW LHRF power and 13.6 MW ICRF power is used to provide a robust, steady state core plasma far from disruptive limits. ARC uses an all-liquid blanket, consisting of low pressure, slowly flowing Fluorine Lithium Beryllium (FLiBe) molten salt. The liquid blanket acts as a working fluid, coolant, and tritium breeder, and minimizes the solid material that can become activated. The large temperature range over which FLiBe is liquid permits blanket operation at 800-900 K with single phase fluid cooling and allows use of a high-efficiency Brayton cycle for electricity production in the secondary coolant loop.

The development of a universal diagnostic probe system for Tokamak fusion test reactor

NASA Technical Reports Server (NTRS)

Mastronardi, R.; Cabral, R.; Manos, D.

1982-01-01

The Tokamak Fusion Test Reactor (TFTR), the largest such facility in the U.S., is discussed with respect to instrumentation in general and mechanisms in particular. The design philosophy and detailed implementation of a universal probe mechanism for TFTR is discussed.

Overview of Alcator C-Mod Research

NASA Astrophysics Data System (ADS)

White, A. E.

2017-10-01

Alcator C-Mod, a compact (R =0.68m, a =0.21m), high magnetic field, Bt <= 8T, tokamak accesses a variety of naturally ELM-suppressed high confinement regimes that feature extreme power density into the divertor, q|| <= 3 GW/m2, with SOL heat flux widths λq <0.5mm, exceeding conditions expected in ITER and approaching those foreseen in power plants. The unique parameter range provides much of the physics basis of a high-field, compact tokamak reactor. Research spans the topics of core transport and turbulence, RF heating and current drive, pedestal physics, scrape-off layer, divertor and plasma wall interactions. In the last experimental campaign, Super H-mode was explored and featured the highest pedestal pressures ever recorded, pped 90 kPa (90% of ITER target), consistent with EPED predictions. Optimization of naturally ELM-suppressed EDA H-modes accessed the highest volume averaged pressures ever achieved (〈p〉>2 atm), with pped 60 kPa. The SOL heat flux width has been measured at Bpol = 1.25T, confirming the Eich scaling over a broader poloidal field range than before. Multi-channel transport studies focus on the relationship between momentum transport and heat transport with perturbative experiments and new multi-scale gyrokinetic simulation validation techniques were developed. U.S. Department of Energy Grant No. DE-FC02-99ER54512.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stencel, J.R.; Finley, V.L.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory for CY90. The report is prepared to provide the US Department of Energy (DOE) and the public with information on the level of radioactive and nonradioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs. The objective of the Annual Site Environmental Report is to document evidence that DOE facility environmental protection programs adequately protect the environment and the public health. The PPPL has engaged in fusion energy research sincemore » 1951 and in 1990 had one of its two large tokamak devices in operation: namely, the Tokamak Fusion Test Reactor. The Princeton Beta Experiment-Modification is undergoing new modifications and upgrades for future operation. A new machine, the Burning Plasma Experiment -- formerly called the Compact Ignition Tokamak -- is under conceptual design, and it is awaiting the approval of its draft Environmental Assessment report by DOE Headquarters. This report is required under the National Environmental Policy Act. The long-range goal of the US Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. 59 refs., 39 figs., 45 tabs.« less

Plasma Physics Lab and the Tokamak Fusion Test Reactor, 1989

ScienceCinema

None

2018-01-16

From the Princeton University Archives: Promotional video about the Plasma Physics Lab and the new Tokamak Fusion Test Reactor (TFTR), with footage of the interior, machines, and scientists at work. This film is discussed in the audiovisual blog of the Seeley G. Mudd Manuscript Library, which holds the archives of Princeton University.

Who will save the tokamak - Harry Potter, Arnold Schwarzenegger, or Shaquille O'Neil?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Mangiarotti, F.; Minervini, J.

2014-10-01

The tokamak is the current leading contender for a fusion power reactor. The reason for the preeminence of the tokamak is its high quality plasma physics performance relative to other concepts. Even so, it is well known that the tokamak must still overcome two basic physics challenges before becoming viable as a DEMO and ultimately a reactor: (1) the achievement of non-inductive steady state operation, and (2) the achievement of robust disruption free operation. These are in addition to the PMI problems faced by all concepts. The work presented here demonstrates by means of a simple but highly credible analytic calculation that a ``standard'' tokamak cannot lead to a reactor - it is just not possible to simultaneously satisfy all the plasma physics plus engineering constraints. Three possible solutions, some more well-known than others, to the problem are analyzed. These visual image generating solutions are defined as (1) the Harry Potter solution, (2) the Arnold Schwarzenegger solution, and (3) the Shaquille O'Neil solution. Each solution will be described both qualitatively and quantitatively at the meeting.

Recent Progress on Spherical Torus Research

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ono, Masayuki; Kaita, Robert

2014-01-01

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A = R0/a) reduced to A ~ 1.5, well below the normal tokamak operating range of A ≥ 2.5. As the aspect ratio is reduced, the ideal tokamak beta β (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as β ~ 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural elongation κ, which makes its plasma shape appear spherical, the ST configurationmore » can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to its longer term goal of attractive fusion energy power source. Since the start of the two megaampere class ST facilities in 2000, National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all of fusion science areas, involving fundamental fusion energy science as well as innovation. These results suggest exciting future prospects for ST research both near term and longer term. The present paper reviews the scientific progress made by the worldwide ST research community during this new mega-ampere-ST era.« less

Compact Torus Injection Experiments on the H.I.T. teststand and the JFT-2M tokamak

NASA Astrophysics Data System (ADS)

Fukumoto, Naoyuki; Fujiwara, Makoto; Kuramoto, Keiji; Ageishi, Masaya; Nagata, Masayoshi; Uyama, Tadao; Ogawa, Hiroaki; Kasai, Satoshi; Hasegawa, Kouichi; Shibata, Takatoshi

1997-11-01

A spheromak-type compact torus (CT) acceleration and injection experiment has been carried out using the Himeji Institute of Technology Compact Torus Injector (HIT-CTI). We investigate the possibility of refueling, density control, current drive, and edge electric field control of tokamak plasmas by means of CT injection. The HIT-CTI produces a CT with a speed of 200 km/s and a density of 1× 10^21m-3. We have constructed new electrodes and power supplies, and will install the HIT-CTI on the JFT-2M tokamak at JAERI in Autumn 1997. The outer electrode serves as a common ground for both the formation bank (144μF, 20kV) and the acceleration bank (92.4μF, 40kV). If the external toroidal field of the tokamak is applied across the CT acceleration region, the CT kinetic energy might decrease during penetration into the field lines joining the inner and outer electrode. This could result in the CT not being able to reach the core of the tokamak plasma. Determining the optimum position of the inner electrode is one of the near term goals of this research. We will present magnetic probe, He-Ne interferometer and fast framing camera data from experiments at H.I.T., where a CT was accelerated into a transverse field. We will also present initial results from the operation of the HIT-CTI on the JFT-2M tokamak.

Source-to-incident-flux relation in a Tokamak blanket module

NASA Astrophysics Data System (ADS)

Imel, G. R.

The next-generation Tokamak experiments, including the Tokamak fusion test reactor (TFTR), will utilize small blanket modules to measure performance parameters such as tritium breeding profiles, power deposition profiles, and neutron flux profiles. Specifically, a neutron calorimeter (simply a neutron moderating blanket module) which permits inferring the incident 14 MeV flux based on measured temperature profiles was proposed for TFTR. The problem of how to relate this total scalar flux to the fusion neutron source is addressed. This relation is necessary since the calorimeter is proposed as a total fusion energy monitor. The methods and assumptions presented was valid for the TFTR Lithium Breeding Module (LBM), as well as other modules on larger Tokamak reactors.

Conceptual design study of the moderate size superconducting spherical tokamak power plant

NASA Astrophysics Data System (ADS)

Gi, Keii; Ono, Yasushi; Nakamura, Makoto; Someya, Youji; Utoh, Hiroyasu; Tobita, Kenji; Ono, Masayuki

2015-06-01

A new conceptual design of the superconducting spherical tokamak (ST) power plant was proposed as an attractive choice for tokamak fusion reactors. We reassessed a possibility of the ST as a power plant using the conservative reactor engineering constraints often used for the conventional tokamak reactor design. An extensive parameters scan which covers all ranges of feasible superconducting ST reactors was completed, and five constraints which include already achieved plasma magnetohydrodynamic (MHD) and confinement parameters in ST experiments were established for the purpose of choosing the optimum operation point. Based on comparison with the estimated future energy costs of electricity (COEs) in Japan, cost-effective ST reactors can be designed if their COEs are smaller than 120 mills kW-1 h-1 (2013). We selected the optimized design point: A = 2.0 and Rp = 5.4 m after considering the maintenance scheme and TF ripple. A self-consistent free-boundary MHD equilibrium and poloidal field coil configuration of the ST reactor were designed by modifying the neutral beam injection system and plasma profiles. The MHD stability of the equilibrium was analysed and a ramp-up scenario was considered for ensuring the new ST design. The optimized moderate-size ST power plant conceptual design realizes realistic plasma and fusion engineering parameters keeping its economic competitiveness against existing energy sources in Japan.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

NASA Astrophysics Data System (ADS)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-01

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritium allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.

Compact Torus Fueling of the STOR-M Tokamak

NASA Astrophysics Data System (ADS)

Xiao, C.; Hirose, A.; Zawalski, W.; White, D.; Raman, R.; Decoste, R.; Gregory, B. C.; Martin, F.

1996-11-01

Tangential injection of accelerated compact torus (CT) has been performed on the STOR-M tokamak (R/a=46/12 cm, B_t<1 T, I_p<= 50 kA, barn_e=(0.5 - 1)×10^13 cm-3) using the University of Saskatchewan Compact Torus Injector (USCTI). The CT parameters are: m~=1 μg, v=120 km/sec, B=0.1 T and n=(2 - 4)×10^15 cm-3. After CT injection, the electron density in tokamak doubles and the poloidal β-value increases. Indications of reduction in the loop voltage and H_α emission level have also been observed. Currently, following efforts are being made: (a) to coat chromium on the electrode surface, (b) to increase the on-line baking temperature, and (c) to reduce the neutral gas load which follows the CT plasma. In addition, numerical calculation of CT motion in a tokamak magnetic field has been carried out. For horizontal injection, the initial CT magnetic dipole direction should be aligned with the CT velocity for deeper penetration. In the case of vertical injection, the CT trajectory is independent of the initial magnetic dipole direction and central penetration is facilitated by off-axis injection.

User's manual for COAST 4: a code for costing and sizing tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sink, D. A.; Iwinski, E. M.

1979-09-01

The purpose of this report is to document the computer program COAST 4 for the user/analyst. COAST, COst And Size Tokamak reactors, provides complete and self-consistent size models for the engineering features of D-T burning tokamak reactors and associated facilities involving a continuum of performance including highly beam driven through ignited plasma devices. TNS (The Next Step) devices with no tritium breeding or electrical power production are handled as well as power producing and fissile producing fusion-fission hybrid reactors. The code has been normalized with a TFTR calculation which is consistent with cost, size, and performance data published in themore » conceptual design report for that device. Information on code development, computer implementation and detailed user instructions are included in the text.« less

Study of neutron generation in the compact tokamak TUMAN-3M in support of a tokamak-based fusion neutron source

NASA Astrophysics Data System (ADS)

Kornev, V. A.; Askinazi, L. G.; Belokurov, A. A.; Chernyshev, F. V.; Lebedev, S. V.; Melnik, A. D.; Shabelsky, A. A.; Tukachinsky, A. S.; Zhubr, N. A.

2017-12-01

The paper presents DD neutron flux measurements in neutron beam injection (NBI) experiments aimed at the optimization of target plasma and heating beam parameters to achieve maximum neutron flux in the TUMAN-3M compact tokamak. Two ion sources of different design were used, which allowed the separation of the beam’s energy and power influence on the neutron rate. Using the database of experiments performed with the two ion sources, an empirical scaling was derived describing the neutron rate dependence on the target plasma and heating beam parameters. Numerical modeling of the neutron rate in the NBI experiments performed using the ASTRA transport code showed good agreement with the scaling.

A fission-fusion hybrid reactor in steady-state L-mode tokamak configuration with natural uranium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reed, Mark; Parker, Ronald R.; Forget, Benoit

2012-06-19

This work develops a conceptual design for a fusion-fission hybrid reactor operating in steady-state L-mode tokamak configuration with a subcritical natural or depleted uranium pebble bed blanket. A liquid lithium-lead alloy breeds enough tritium to replenish that consumed by the D-T fusion reaction. The fission blanket augments the fusion power such that the fusion core itself need not have a high power gain, thus allowing for fully non-inductive (steady-state) low confinement mode (L-mode) operation at relatively small physical dimensions. A neutron transport Monte Carlo code models the natural uranium fission blanket. Maximizing the fission power gain while breeding sufficient tritiummore » allows for the selection of an optimal set of blanket parameters, which yields a maximum prudent fission power gain of approximately 7. A 0-D tokamak model suffices to analyze approximate tokamak operating conditions. This fission blanket would allow the fusion component of a hybrid reactor with the same dimensions as ITER to operate in steady-state L-mode very comfortably with a fusion power gain of 6.7 and a thermal fusion power of 2.1 GW. Taking this further can determine the approximate minimum scale for a steady-state L-mode tokamak hybrid reactor, which is a major radius of 5.2 m and an aspect ratio of 2.8. This minimum scale device operates barely within the steady-state L-mode realm with a thermal fusion power of 1.7 GW. Basic thermal hydraulic analysis demonstrates that pressurized helium could cool the pebble bed fission blanket with a flow rate below 10 m/s. The Brayton cycle thermal efficiency is 41%. This reactor, dubbed the Steady-state L-mode non-Enriched Uranium Tokamak Hybrid (SLEUTH), with its very fast neutron spectrum, could be superior to pure fission reactors in terms of breeding fissile fuel and transmuting deleterious fission products. It would likely function best as a prolific plutonium breeder, and the plutonium it produces could actually be more proliferation-resistant than that bred by conventional fast reactors. Furthermore, it can maintain constant total hybrid power output as burnup proceeds by varying the neutron source strength.« less

Recent Progress on Spherical Torus Research and Implications for Fusion Energy Development Path

NASA Astrophysics Data System (ADS)

Ono, Masayuki

2014-10-01

The spherical torus or spherical tokamak (ST) is a member of the tokamak family with its aspect ratio (A =R0 / a) reduced to A near 1.5, well below the normal tokamak operating range of A equal to 2.5 or greater. As the aspect ratio is reduced, the ideal tokamak beta (radio of plasma to magnetic pressure) stability limit increases rapidly, approximately as 1/A. The plasma current it can sustain for a given edge safety factor q-95 also increases rapidly. Because of the above, as well as the natural plasma elongation which makes its plasma shape appear spherical, the ST configuration can yield exceptionally high tokamak performance in a compact geometry. Due to its compactness and high performance, the ST configuration has various near term applications, including a compact fusion neutron source with low tritium consumption, in addition to the longer term goal of an attractive fusion energy power source. Since the start of the two mega-ampere class ST facilities in 2000, the National Spherical Torus Experiment (NSTX) in the US and Mega Ampere Spherical Tokamak (MAST) in the UK, active ST research has been conducted worldwide. More than sixteen ST research facilities operating during this period have achieved remarkable advances in all areas of fusion research, including fundamental fusion energy science as well as technological innovation. These results suggest exciting future prospects for ST research in both the near and longer term. The talk will summarize the key physics results from worldwide ST experiments, and describe ST community plans to provide the database for FNSF design while improving predictive capabilities for ITER and beyond. This work supported by DoE Contract No. DE-AC02-09CH11466.

Georgia Tech Studies of Sub-Critical Advanced Burner Reactors with a D-T Fusion Tokamak Neutron Source for the Transmutation of Spent Nuclear Fuel

NASA Astrophysics Data System (ADS)

Stacey, W. M.

2009-09-01

The possibility that a tokamak D-T fusion neutron source, based on ITER physics and technology, could be used to drive sub-critical, fast-spectrum nuclear reactors fueled with the transuranics (TRU) in spent nuclear fuel discharged from conventional nuclear reactors has been investigated at Georgia Tech in a series of studies which are summarized in this paper. It is found that sub-critical operation of such fast transmutation reactors is advantageous in allowing longer fuel residence time, hence greater TRU burnup between fuel reprocessing stages, and in allowing higher TRU loading without compromising safety, relative to what could be achieved in a similar critical transmutation reactor. The required plasma and fusion technology operating parameter range of the fusion neutron source is generally within the anticipated operational range of ITER. The implications of these results for fusion development policy, if they hold up under more extensive and detailed analysis, is that a D-T fusion tokamak neutron source for a sub-critical transmutation reactor, built on the basis of the ITER operating experience, could possibly be a logical next step after ITER on the path to fusion electrical power reactors. At the same time, such an application would allow fusion to contribute to meeting the nation's energy needs at an earlier stage by helping to close the fission reactor nuclear fuel cycle.

Nuclear design of a very-low-activation fusion reactor

NASA Astrophysics Data System (ADS)

Cheng, E. T.; Hopkins, G. R.

1983-06-01

The nuclear design aspects of using very-low-activation materials, such as SiC, MgO, and aluminum for fusion-reactor first wall, blanket, and shield applications were investigated. In addition to the advantage of very-low radioactive inventory, it was found that the very-low-activation fusion reactor can also offer an adequate tritium-breeding ratio and substantial amount of blanket nuclear heating as a conventional-material-structured reactor does. The most-stringent design constraint found in a very-low-activation fusion reactor is the limited space available in the inboard region of a Tokamak concept for shielding to protect the superconducting toroidal field coil. A reference design was developed which mitigates the constraint by adopting a removable tungsten shield design that retains the inboard dimensions and gives the same shield performance as the reference STARFIRE Tokamak reactor design.

Scoping and sensitivity analyses for the Demonstration Tokamak Hybrid Reactor (DTHR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sink, D.A.; Gibson, G.

1979-03-01

The results of an extensive set of parametric studies are presented which provide analytical data of the effects of various tokamak parameters on the performance and cost of the DTHR (Demonstration Tokamak Hybrid Reactor). The studies were centered on a point design which is described in detail. Variations in the device size, neutron wall loading, and plasma aspect ratio are presented, and the effects on direct hardware costs, fissile fuel production (breeding), fusion power production, electrical power consumption, and thermal power production are shown graphically. The studies considered both ignition and beam-driven operations of DTHR and yielded results based onmore » two empirical scaling laws presently used in reactor studies. Sensitivity studies were also made for variations in the following key parameters: the plasma elongation, the minor radius, the TF coil peak field, the neutral beam injection power, and the Z/sub eff/ of the plasma.« less

Solenoid-free plasma start-up in spherical tokamaks

NASA Astrophysics Data System (ADS)

Raman, R.; Shevchenko, V. F.

2014-10-01

The central solenoid is an intrinsic part of all present-day tokamaks and most spherical tokamaks. The spherical torus (ST) confinement concept is projected to operate at high toroidal beta and at a high fraction of the non-inductive bootstrap current as required for an efficient reactor system. The use of a conventional solenoid in a ST-based fusion nuclear facility is generally believed to not be a possibility. Solenoid-free plasma start-up is therefore an area of extensive worldwide research activity. Solenoid-free plasma start-up is also relevant to steady-state tokamak operation, as the central transformer coil of a conventional aspect ratio tokamak reactor would be located in a high radiation environment but would be needed only during the initial discharge initiation and current ramp-up phases. Solenoid-free operation also provides greater flexibility in the selection of the aspect ratio and simplifies the reactor design. Plasma start-up methods based on induction from external poloidal field coils, helicity injection and radio frequency current drive have all made substantial progress towards meeting this important need for the ST. Some of these systems will now undergo the final stages of test in a new generation of large STs, which are scheduled to begin operations during the next two years. This paper reviews research to date on methods for inducing the initial start-up current in STs without reliance on the conventional central solenoid.

Plasma-material Interactions in Current Tokamaks and their Implications for Next-step Fusion Reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Federici, G.; Skinner, C.H.; Brooks, J.N.

2001-01-10

The major increase in discharge duration and plasma energy in a next-step DT [deuterium-tritium] fusion reactor will give rise to important plasma-material effects that will critically influence its operation, safety, and performance. Erosion will increase to a scale of several centimeters from being barely measurable at a micron scale in today's tokamaks. Tritium co-deposited with carbon will strongly affect the operation of machines with carbon plasma-facing components. Controlling plasma wall interactions is critical to achieving high performance in present-day tokamaks and this is likely to continue to be the case in the approach to practical fusion reactors. Recognition of themore » important consequences of these phenomena has stimulated an internationally coordinated effort in the field of plasma-surface interactions supporting the Engineering Design Activities of the International Thermonuclear Experimental Reactor (ITER) project and significant progress has been made in better under standing these issues. This paper reviews the underlying physical processes and the existing experimental database of plasma-material interactions both in tokamaks and laboratory simulation facilities for conditions of direct relevance to next-step fusion reactors. Two main topical groups of interactions are considered: (i) erosion/redeposition from plasma sputtering and disruptions, including dust and flake generation, (ii) tritium retention and removal. The use of modeling tools to interpret the experimental results and make projections for conditions expected in future devices is explained. Outstanding technical issues and specific recommendations on potential R and D [Research and Development] avenues for their resolution are presented.« less

Design of the DEMO Fusion Reactor Following ITER.

PubMed

Garabedian, Paul R; McFadden, Geoffrey B

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task.

Design of the DEMO Fusion Reactor Following ITER

PubMed Central

Garabedian, Paul R.; McFadden, Geoffrey B.

2009-01-01

Runs of the NSTAB nonlinear stability code show there are many three-dimensional (3D) solutions of the advanced tokamak problem subject to axially symmetric boundary conditions. These numerical simulations based on mathematical equations in conservation form predict that the ITER international tokamak project will encounter persistent disruptions and edge localized mode (ELMS) crashes. Test particle runs of the TRAN transport code suggest that for quasineutrality to prevail in tokamaks a certain minimum level of 3D asymmetry of the magnetic spectrum is required which is comparable to that found in quasiaxially symmetric (QAS) stellarators. The computational theory suggests that a QAS stellarator with two field periods and proportions like those of ITER is a good candidate for a fusion reactor. For a demonstration reactor (DEMO) we seek an experiment that combines the best features of ITER, with a system of QAS coils providing external rotational transform, which is a measure of the poloidal field. We have discovered a configuration with unusually good quasisymmetry that is ideal for this task. PMID:27504224

Reflux physics and an operational scenario for the spheromak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hooper, E. B.

2010-07-20

The spheromak [1] is a toroidal magnetic confinement geometry for plasma with most of the magnetic field generated by internal currents. It has been demonstrated to have excellent energy confinement properties: A peak electron temperature of 0.4 keV was achieved in the Compact Torus Experiment (CTX) experiment [2] and of 0.5 keV in the Sustained Spheromak Physics Experiment (SSPX) [3]. In both cases the plasmas were decaying slowly following formation and (in SSPX) sustainment by coaxial helicity injection (CHI) [4]. In SSPX, power balance analysis during this operational phase yielded electron thermal conductivities in the core plasma in the rangemore » of 1-10 m 2/s [5, 6], comparable to the tokamak L-mode. These results motivate the consideration of possible operating scenarios for future fusion experiments or even reactors.« less

Fusion Breeding for Sustainable, Mid Century, Carbon Free Power

NASA Astrophysics Data System (ADS)

Manheimer, Wallace

2015-11-01

If ITER achieves Q ~10, it is still very far from useful fusion. The fusion power, and the driver power will allow only a small amount of power to be delivered, <~50MW for an ITER scale tokamak. It is unlikely, considering ``conservative design rules'' that tokamaks can ever be economical pure fusion power producers. Considering the status of other magnetic fusion concepts, it is also very unlikely that any alternate concept will either. Laser fusion does not seem to be constrained by any conservative design rules, but considering the failure of NIF to achhieve ignition, at this point it has many more obstacles to overcome than magnetic fusion. One way out of this dilemma is to use an ITER size tokamak, or a NIF size laser, as a fuel breeder for searate nuclear reactors. Hence ITER and NIF become ends in themselves, instead of steps to who knows what DEMO decades later. Such a tokamak can easily live within the consrtaints of conservative design rules. This has led the author to propose ``The Energy Park'' a sustainable, carbon free, economical, and environmently viable power source without prolifertion risk. It is one fusion breeder fuels 5 conventional nuclear reactors, and one fast neutron reactor burns the actinide wastes.

GEM detector development for tokamak plasma radiation diagnostics: SXR poloidal tomography

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Malinowski, Karol; Ziółkowski, Adam; Kowalska-Strzeciwilk, Ewa; Czarski, Tomasz; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Kolasiński, Piotr; Krawczyk, Rafał D.

2015-09-01

An increased attention to tungsten material is related to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. The proposed work refers to the studies of W influence on the plasma performances by developing new detectors based on Gas Electron Multiplier GEM) technology for tomographic studies of tungsten transport in ITER-oriented tokamaks, e.g. WEST project. It presents current stage of design and developing of cylindrically bent SXR GEM detector construction for horizontal port implementation. Concept to overcome an influence of constraints on vertical port has been also presented. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing creation of sustainable nuclear fusion reactors a step closer.

Apollo - An advanced fuel fusion power reactor for the 21st century

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulcinski, G.L.; Emmert, G.A.; Blanchard, J.P.

1989-03-01

A preconceptual design of a tokamak reactor fueled by a D-He-3 plasma is presented. A low aspect ratio (A=2-4) device is studied here but high aspect ratio devices (A > 6) may also be quite attractive. The Apollo D-He-3 tokamak capitalizes on recent advances in high field magnets (20 T) and utilizes rectennas to convert the synchrotron radiation directly to electricity. The overall efficiency ranges from 37 to 52% depending on whether the bremsstrahlung energy is utilized. The low neutron wall loading (0.1 MW/m/sup 2/) allows a permanent first wall to be designed and the low nuclear decay heat enablesmore » the reactor to be classed as inherently safe. The cost of electricity from Apollo is > 40% lower than electricity from a similar sized DT reactor.« less

Stainless steel blanket concept for tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

Karbowski, J.S.; Lee, A.Y.; Prevenslik, T.V.

1979-01-25

The purpose of this joint ORNL/Westinghouse Program is to develop a design concept for a tokamak reactor blanket system which satisfies engineering requirements for a utility environment. While previous blanket studies have focused primarily on performance issues (thermal, neutronic, and structural), this study has emphasized consideration of reliability, fabricability, and lifetime.

Physics evaluation of compact tokamak ignition experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uckan, N.A.; Houlberg, W.A.; Sheffield, J.

1985-01-01

At present, several approaches for compact, high-field tokamak ignition experiments are being considered. A comprehensive method for analyzing the potential physics operating regimes and plasma performance characteristics of such ignition experiments with O-D (analytic) and 1-1/2-D (WHIST) transport models is presented. The results from both calculations are in agreement and show that there are regimes in parameter space in which a class of small (R/sub o/ approx. 1-2 m), high-field (B/sub o/ approx. 8-13 T) tokamaks with aB/sub o/S/q/sub */ approx. 25 +- 5 and kappa = b/a approx. 1.6-2.0 appears ignitable for a reasonable range of transport assumptions. Consideringmore » both the density and beta limits, an evaluation of the performance is presented for various forms of chi/sub e/ and chi/sub i/, including degradation at high power and sawtooth activity. The prospects of ohmic ignition are also examined. 16 refs., 13 figs.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high- field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and densitymore » regime consistent with a hot-ion-mode fusion reactor. As a result, these simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high-field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and density regimemore » consistent with a hot-ion-mode fusion reactor. These simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for 233U fuel production: 1. Concept of a stationary Tokamak as a neutron source

NASA Astrophysics Data System (ADS)

Azizov, E. A.; Gladush, G. G.; Dokuka, V. N.; Khayrutdinov, R. R.

2015-12-01

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of 233U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket based on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.

A high speed compact microwave interferometer for density fluctuation measurements in Sino-UNIted Spherical Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhong, H., E-mail: zhongh14@126.com; Tan, Y.; Liu, Y. Q.

2016-11-15

A single-channel 3 mm interferometer has been developed for plasma density diagnostics in the Sino-UNIted Spherical Tokamak (SUNIST). The extremely compact microwave interferometer utilizes one corrugated feed horn antenna for both emitting and receiving the microwave. The beam path lies on the equatorial plane so the system would not suffer from beam path deflection problems due to the symmetry of the cross section. A focusing lens group and an oblique vacuum window are carefully designed to boost the signal to noise ratio, which allows this system to show good performance even with the small-diameter central column itself as a reflector,more » without a concave mirror. The whole system discards the reference leg for maximum compactness, which is particularly suitable for the small-sized tokamak. An auto-correcting algorithm is developed to calculate the phase evolution, and the result displays good phase stability of the whole system. The intermediate frequency is adjustable and can reach its full potential of 2 MHz for best temporal resolution. Multiple measurements during ohmic discharges proved the interferometer’s capability to track typical density fluctuations in SUNIST, which enables this system to be utilized in the study of MHD activities.« less

Introduction to D-He(3) fusion reactors

NASA Technical Reports Server (NTRS)

Vlases, G. C.; Steinhauer, L. C.

1989-01-01

A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

Tokamak reactor for treating fertile material or waste nuclear by-products

DOEpatents

Kotschenreuther, Michael T.; Mahajan, Swadesh M.; Valanju, Prashant M.

2012-10-02

Disclosed is a tokamak reactor. The reactor includes a first toroidal chamber, current carrying conductors, at least one divertor plate within the first toroidal chamber and a second chamber adjacent to the first toroidal chamber surrounded by a section that insulates the reactor from neutrons. The current carrying conductors are configured to confine a core plasma within enclosed walls of the first toroidal chamber such that the core plasma has an elongation of 1.5 to 4 and produce within the first toroidal chamber at least one stagnation point at a perpendicular distance from an equatorial plane through the core plasma that is greater than the plasma minor radius. The at least one divertor plate and current carrying conductors are configured relative to one another such that the current carrying conductors expand the open magnetic field lines at the divertor plate.

Introduction to D-He(3) fusion reactors

NASA Astrophysics Data System (ADS)

Vlases, G. C.; Steinhauer, L. C.

1989-07-01

A review and evaluation of D-He(3) fusion reactor technology is presented. The advantages and disadvantages of the D-He(3) and D-T reactor cycles are outlined and compared. In addition, the general design features of D-He(3) tokamaks and field reversed configuration (FRC) reactors are described and the relative merits of each are compared. It is concluded that both tokamaks and FRC's offer certain advantages, and that the ultimate decision as to which to persue for terrestrial power generation will depend heavily on how the physics performance of each of them develops over the next few years. It is clear that the D-He(3) fuel cycle offers marked advantages over the D-T cycle. Although the physics requirements for D-He(3) are more demanding, the overwhelming advantages resulting from the two order of magnitude reduction of neutron flux are expected to lead to a shorter time to commercialization than for the D-T cycle.

Alternate fusion fuels workshop

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1981-06-01

The workshop was organized to focus on a specific confinement scheme: the tokamak. The workshop was divided into two parts: systems and physics. The topics discussed in the systems session were narrowly focused on systems and engineering considerations in the tokamak geometry. The workshop participants reviewed the status of system studies, trade-offs between d-t and d-d based reactors and engineering problems associated with the design of a high-temperature, high-field reactor utilizing advanced fuels. In the physics session issues were discussed dealing with high-beta stability, synchrotron losses and transport in alternate fuel systems. The agenda for the workshop is attached.

Integrated Tokamak modeling: When physics informs engineering and research planning

NASA Astrophysics Data System (ADS)

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. It discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Poli, Francesca Maria

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

Integrated Tokamak modeling: When physics informs engineering and research planning

DOE PAGES

Poli, Francesca Maria

2018-05-01

Modeling tokamaks enables a deeper understanding of how to run and control our experiments and how to design stable and reliable reactors. We model tokamaks to understand the nonlinear dynamics of plasmas embedded in magnetic fields and contained by finite size, conducting structures, and the interplay between turbulence, magneto-hydrodynamic instabilities, and wave propagation. This tutorial guides through the components of a tokamak simulator, highlighting how high-fidelity simulations can guide the development of reduced models that can be used to understand how the dynamics at a small scale and short time scales affects macroscopic transport and global stability of plasmas. Itmore » discusses the important role that reduced models have in the modeling of an entire plasma discharge from startup to termination, the limits of these models, and how they can be improved. It discusses the important role that efficient workflows have in the coupling between codes, in the validation of models against experiments and in the verification of theoretical models. Finally, it reviews the status of integrated modeling and addresses the gaps and needs towards predictions of future devices and fusion reactors.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Redi, M.H.; Mynick, H.E.; Suewattana, M.

Hamiltonian coordinate, guiding-center code calculations of the confinement of suprathermal ions in quasi-axisymmetric stellarator (QAS) designs have been carried out to evaluate the attractiveness of compact configurations which are optimized for ballooning stability. A new stellarator particle-following code is used to predict ion loss rates and particle confinement for thermal and neutral beam ions in a small experiment with R = 145 cm, B = 1-2 T and for alpha particles in a reactor-size device. In contrast to tokamaks, it is found that high edge poloidal flux has limited value in improving ion confinement in QAS, since collisional pitch-angle scatteringmore » drives ions into ripple wells and stochastic field regions, where they are quickly lost. The necessity for reduced stellarator ripple fields is emphasized. The high neutral beam ion loss predicted for these configurations suggests that more interesting physics could be explored with an experiment of less constrained size and magnetic field geometry.« less

Li Experiments at the Tokamak T-11M Toward PFC Concept of Steady State Tokamak-Reactor

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2009-11-01

As practical method of using a liquid lithium as a renewable plasma-facing component (PCF) for steady state tokamak-reactor the concept of lithium emitter-collector is considered [1]. It is based on lithium filled capillary porous system proposed by V.A. Evtikhin et al. (1996). The lithium circulation process consists of four steps: (1) Li emission from the PFC emitter into the plasma; (2) plasma boundary cooling by non-coronal Li radiation; (3) Li ion capture by the collector (before they are lost to the tokamak chamber wall); (4) Li return from the collector to the emitter. T-11M tokamak experiments have used three local rail limiters made from lithium, molybdenum and graphite as lithium collectors. The lithium behavior was studied by analysis of the witness samples, and by a mobile graphite probe. The key findings are: (1) lithium collection on the ion side of the lithium limiter is 2-3 times larger than on the electron side; (2) total efficiency of Li collection integrated over all three rail limiters can reach 50-70% of the lithium emission during the discharge pulse, while the theoretical limit is about 90%. [1] S.V. Mirnov, J. Nucl. Mat., 390-391, 876 (2009).

Influence of fast alpha diffusion and thermal alpha buildup on tokamak reactor performance

DOE Office of Scientific and Technical Information (OSTI.GOV)

Uckan, N.A.; Tolliver, J.S.; Houlberg, W.A.

1987-11-01

The effect of fast alpha diffusion and thermal alpha accumulation on the confinement capability of a candidate Engineering Test Reactor (ETR) plasma (Tokamak Ignition/Burn Experimental Reactor (TIBER-II)) in achieving ignition and steady-state driven operation has been assessed using both global and 1-1/2-D transport models. Estimates are made of the threshold for radial diffusion of fast alphas and thermal alpha buildup. It is shown that a relatively low level of radial transport, when combined with large gradients in the fast alpha density, leads to a significant radial flow with a deleterious effect on plasma performance. Similarly, modest levels of thermal alphamore » concentration significantly influence the ignition and steady-state burn capability. 23 refs., 9 figs., 4 tabs.« less

Alpha channeling with high-field launch of lower hybrid waves

DOE PAGES

Ochs, I. E.; Bertelli, N.; Fisch, N. J.

2015-11-04

Although lower hybrid waves are effective at driving currents in present-day tokamaks, they are expected to interact strongly with high-energy particles in extrapolating to reactors. In the presence of a radial alpha particle birth gradient, this interaction can take the form of wave amplification rather than damping. While it is known that this amplification more easily occurs when launching from the tokamak high-field side, the extent of this amplification has not been made quantitative. Here, by tracing rays launched from the high- field-side of a tokamak, the required radial gradients to achieve amplification are calculated for a temperature and densitymore » regime consistent with a hot-ion-mode fusion reactor. As a result, these simulations, while valid only in the linear regime of wave amplification, nonetheless illustrate the possibilities for wave amplification using high-field launch of the lower hybrid wave.« less

Concept of a demonstrational hybrid reactor—a tokamak with molten-salt blanket for {sup 233}U fuel production: 1. Concept of a stationary Tokamak as a neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Azizov, E. A.; Gladush, G. G., E-mail: gladush@triniti.ru; Dokuka, V. N.

2015-12-15

On the basis of current understanding of physical processes in tokamaks and taking into account engineering constraints, it is shown that a low-cost facility of a moderate size can be designed within the adopted concept. This facility makes it possible to achieve the power density of neutron flux which is of interest, in particular, for solving the problem of {sup 233}U fuel production from thorium. By using a molten-salt blanket, the important task of ensuring the safe operation of such a reactor in the case of possible coolant loss is accomplished. Moreover, in a hybrid reactor with the blanket basedmore » on liquid salts, the problem of periodic refueling that is difficult to perform in solid blankets can be solved.« less

Can tokamaks PFC survive a single event of any plasma instabilities?

NASA Astrophysics Data System (ADS)

Hassanein, A.; Sizyuk, V.; Miloshevsky, G.; Sizyuk, T.

2013-07-01

Plasma instability events such as disruptions, edge-localized modes (ELMs), runaway electrons (REs), and vertical displacement events (VDEs) are continued to be serious events and most limiting factors for successful tokamak reactor concept. The plasma-facing components (PFCs), e.g., wall, divertor, and limited surfaces of a tokamak as well as coolant structure materials are subjected to intense particle and heat loads and must maintain a clean and stable surface environment among them and the core/edge plasma. Typical ITER transient events parameters are used for assessing the damage from these four different instability events. HEIGHTS simulation showed that a single event of a disruption, giant ELM, VDE, or RE can cause significant surface erosion (melting and vaporization) damage to PFC, nearby components, and/or structural materials (VDE, RE) melting and possible burnout of coolant tubes that could result in shut down of reactor for extended repair time.

Neutral atom analyzers for diagnosing hot plasmas: A review of research at the ioffe physicotechnical institute

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kislyakov, A. I.; Petrov, M. P.

2009-07-15

Research on neutral particle diagnostics of thermonuclear plasmas that has been carried out in recent years at the Ioffe Physicotechnical Institute of the Russian Academy of Sciences (St. Petersburg, Russia) is reviewed. Work on the creation and improvement of neutral atom analyzers was done in two directions: for potential applications (in particular, on the International Thermonuclear Experimental Reactor, which is now under construction at Cadarache in France) and for investigation of the ion plasma component in various devices (in particular, in the largest tokamaks, such as JET, TFTR, and JT-60). Neutral atom analyzers are the main tool for studying themore » behavior of hydrogen ions and isotopes in magnetic confinement systems. They make it possible to determine energy spectra, to perform the isotope analysis of atom fluxes from the plasma, to measure the absolute intensity of the fluxes, and to record how these parameters vary with time. A comparative description of the analyzers developed in recent years at the Ioffe Institute is given. These are ACORD-12/24 analyzers for recording 0.2-100-keV hydrogen and deuterium atoms with a tunable range of simultaneously measured energies, CNPA compact analyzers for a fixed energy gain in the ranges 80-1000 eV and 0.8-100 keV, an ISEP analyzer for simultaneously recording the atoms of all the three hydrogen isotopes (H, D, and T) in the energy range 5-700 keV, and GEMMA analyzers for recording atom fluxes of hydrogen and helium isotopes in the range 0.1-4 MeV. The scintillating detectors of the ISEP and GEMMA analyzers have a lowered sensitivity to neutrons and thus can operate without additional shielding in neutron fields of up to 10{sup 9} n/(cm{sup 2} s). These two types of analyzers, intended to operate under deuterium-tritium plasma conditions, are prototypes of atom analyzers created at the Ioffe Institute for use in the International Thermonuclear Experimental Reactor. With these analyzers, a number of new results have been obtained in recent years in various devices. Some results are presented from investigation of ions in the Globus-M spherical tokamak, the W7-AS stellarator, and the JET tokamak by means of the analyzers developed at the Ioffe Institute. Challenges and opportunities for applying these diagnostics in the International Thermonuclear Experimental Reactor project are discussed.« less

Scaling mechanisms of vapour/plasma shielding from laser-produced plasmas to magnetic fusion regimes

NASA Astrophysics Data System (ADS)

Sizyuk, Tatyana; Hassanein, Ahmed

2014-02-01

The plasma shielding effect is a well-known mechanism in laser-produced plasmas (LPPs) reducing laser photon transmission to the target and, as a result, significantly reducing target heating and erosion. The shielding effect is less pronounced at low laser intensities, when low evaporation rate together with vapour/plasma expansion processes prevent establishment of a dense plasma layer above the surface. Plasma shielding also loses its effectiveness at high laser intensities when the formed hot dense plasma plume causes extensive target erosion due to radiation fluxes back to the surface. The magnitude of emitted radiation fluxes from such a plasma is similar to or slightly higher than the laser photon flux in the low shielding regime. Thus, shielding efficiency in LPPs has a peak that depends on the laser beam parameters and the target material. A similar tendency is also expected in other plasma-operating devices such as tokamaks of magnetic fusion energy (MFE) reactors during transient plasma operation and disruptions on chamber walls when deposition of the high-energy transient plasma can cause severe erosion and damage to the plasma-facing and nearby components. A detailed analysis of these abnormal events and their consequences in future power reactors is limited in current tokamak reactors. Predictions for high-power future tokamaks are possible only through comprehensive, time-consuming and rigorous modelling. We developed scaling mechanisms, based on modelling of LPP devices with their typical temporal and spatial scales, to simulate tokamak abnormal operating regimes to study wall erosion, plasma shielding and radiation under MFE reactor conditions. We found an analogy in regimes and results of carbon and tungsten erosion of the divertor surface in ITER-like reactors with erosion due to laser irradiation. Such an approach will allow utilizing validated modelling combined with well-designed and well-diagnosed LPP experimental studies for predicting consequences of plasma instabilities in complex fusion environment, which are of serious concern for successful energy production.

Twenty Years of Research on the Alcator C-Mod Tokamak

NASA Astrophysics Data System (ADS)

Greenwald, Martin

2013-10-01

Alcator C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since its start in 1993, contributing data that extended tests of critical physical models into new parameter ranges and into new regimes. Using only RF for heating and current drive with innovative launching structures, C-Mod operates routinely at very high power densities. Research highlights include direct experimental observation of ICRF mode-conversion, ICRF flow drive, demonstration of Lower-Hybrid current drive at ITER-like densities and fields and, using a set of powerful new diagnostics, extensive validation of advanced RF codes. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components--an approach adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and EDA H-mode regimes which have high performance without large ELMs and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and found that self-generated flow shear can be strong enough to significantly modify transport. C-Mod made the first quantitative link between pedestal temperature and H-mode performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. Work supported by U.S. DoE

Protection of tokamak plasma facing components by a capillary porous system with lithium

NASA Astrophysics Data System (ADS)

Lyublinski, I.; Vertkov, A.; Mirnov, S.; Lazarev, V.

2015-08-01

Development of plasma facing material (PFM) based on the Capillary-Porous System (CPS) with lithium and activity on realization of lithium application strategy are addressed to meet the challenges under the creation of steady-state tokamak fusion reactor and fusion neutron source. Presented overview of experimental study of lithium CPS in plasma devices demonstrates the progress in protection of tokamak plasma facing components (PFC) from damage, stabilization and self-renewal of liquid lithium surface, elimination of plasma pollution and lithium accumulation in tokamak chamber. The possibility of PFC protection from the high power load related to cooling of the tokamak boundary plasma by radiation of non-fully stripped lithium ions supported by experimental results. This approach demonstrated in scheme of closed loops of Li circulation in the tokamak vacuum chamber and realized in a series of design of tokamak in-vessel elements.

Design of an advanced bundle divertor for the Demonstration Tokamak Hybrid Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, T.F.; Lee, A.Y.; Ruck, G.W.

1979-01-25

The conclusion of this work is that a bundle divertor, using an improved method of designing the magnetic field configuration, is feasible for the Demonstration Tokamak Hybrid Reactor (DTHR) investigated by Westinghouse. The most significant achievement of this design is the reduction in current density (1 kA/cm/sup 2/) in the divertor coils in comparison to the overall averaged current densities per tesla of field to be nulled for DITE (25 kA/cm/sup 2/) and for ISX-B/sup 2/ (11 kA/cm/sup 2/). Therefore, superconducting magnets can be built into the tight space available with a sound mechanical structure.

Pellet injector development at ORNL (Oak Ridge National Laboratory)

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Argo, B. E.; Baylor, L. R.; Combs, S. K.; Fehling, D. T.; Fisher, P. W.; Foster, C. A.; Foust, C. R.; Milora, S. L.; Qualls, A. L.

1990-09-01

Advanced plasma fueling systems for magnetic confinement experiments are under development at Oak Ridge National Laboratory (ORNL). The general approach is that of producing and accelerating frozen hydrogenic pellets to speeds in the kilometer-per-second range by either pneumatic (light-gas gun) or mechanical (centrifugal force) techniques. ORNL has recently provided a centrifugal pellet injector for the Tore Supra tokamak and a new, simplified, eight-shot pneumatic injector for the Advanced Toroidal Facility stellarator at ORNL. Hundreds of tritium and DT pellets were accelerated at the Tritium Systems Test Assembly facility at Los Alamos in 1988 to 1989. These experiments, done in a single-shot pipe-gun system, demonstrated the feasibility of forming and accelerating tritium pellets at low (sup 3)He levels. A new, tritium-compatible extruder mechanism is being designed for longer-pulse DT applications. Two-stage light-gas guns and electron beam rocket accelerators for speeds of the order of 2 to 10 km/s are also under development. Recently, a repeating, two-stage light-gas gun accelerated 10 surrogate pellets at a 1-Hz repetition rate to speeds in the range of 2 to 3 km/s; and the electron beam rocket accelerator completed initial feasibility and scaling experiments. ORNL has also developed conceptual designs of advanced plasma fueling systems for the Compact Ignition Tokamak and the International Thermonuclear Experimental Reactor.

High Performance Regimes in Alcator C-Mod at High Magnetic Field

NASA Astrophysics Data System (ADS)

Marmar, E. S.; Alcator C-Mod Team

2017-10-01

Alcator is the only divertor tokamak in the world capable of operating at magnetic fields up to 8 T, equaling and exceeding that planned for ITER. Using RF and microwave tools for auxiliary heating and current drive, C-Mod accesses high pressure, high density, reactor-relevant regimes with no external torque and equilibrated electrons and ions, with exclusive use of high-Z metal plasma-facing components. The 2016 experimental campaign focused on naturally ELM-suppressed, enhanced energy confinement regimes (including I-mode and EDA H-mode, and approaches to super-H-mode), with emphasis on operation at the highest fields (52 atm.) was achieved. Taken together, combined with previous results from C-Mod and the world tokamak database, these results form a strong foundation for the high field, compact approach to achieving fusion energy production. New advances in high temperature, high field superconductors open the possibilities for practical development of this path for commercial fusion. Supported by USDOE.

From pure fusion to fusion-fission Demo tokamaks

NASA Astrophysics Data System (ADS)

Mirnov, S. V.

2013-04-01

The major requirements for pure fusion tokamak reactors and tokamak-based fusion neutron sources (FNS) are analyzed together with possible paths from the present-day tokamak towards the FNS tokamak. The FNS are of interest for traditional fission reactors as a method of waste management by burning of long-lived transuranic radionuclides (minorities) and fission fuel breeding. The Russian fission community places several hard requirements on the quality of FNS suitable for the first step of the investigation program of minority burning and breeding. They are (a) a steady-state regime of neutron production (more than 80% of the operational time), (b) a neutron power flux density greater than >0.2 MW m-2, (c) a total surface integrated neutron power >10 MW. Among the different FNS projects, based on magnetically confined plasmas, only ‘classical tokamak’ is most likely to fulfill these requirements in the nearest future. Some of the most important improvements of the ‘classical tokamak’ needed for successful realization of the FNS are (1) decrease in Zeff (probably, by making use of lithium as a part of plasma-facing components), (2) He removal and closed loop DT fuel circulation, (3) increase in the energy of stationary injected neutral tritium beams up to 150-170 keV and (4) control of impurity contamination at the plasma center (probably, by local RF heating). These key issues are discussed.

Fast Time Response Electromagnetic Disruption Mitigation Concept

DOE PAGES

Raman, R.; Jarboe, T.; Jernigan, Thomas C.; ...

2015-09-28

An important and urgent issue for ITER is predicting and controlling disruptions. Tokamaks and spherical tokamaks have the potential to disrupt. Methods to rapidly quench the discharge after an impending disruption is detected are essential to protect the vessel and internal components. The warning time for the onset of some disruptions in tokamaks could be <10 ms, which poses stringent requirements on the disruption mitigation system for reactor systems. In this proposed method, a cylindrical boron nitride projectile containing a radiative payload composed of boron, boron nitride, or beryllium particulate matter and weighing similar to 15 g is accelerated tomore » velocities on the order of 1 to 2 km/s in <2 ms in a linear rail gun accelerator. A partially fragmented capsule is then injected into the tokamak discharge in the 3- to 6-ms timescale, where the radiative payload is dispersed. The device referred to as an electromagnetic particle injector has the potential to meet the short warning timescales for which a reactor disruption mitigation system must be built. The system is fully electromagnetic, with no mechanical moving parts, which ensures high reliability after a period of long standby.« less

How much does a tokamak reactor cost?

NASA Astrophysics Data System (ADS)

Freidberg, J.; Cerfon, A.; Ballinger, S.; Barber, J.; Dogra, A.; McCarthy, W.; Milanese, L.; Mouratidis, T.; Redman, W.; Sandberg, A.; Segal, D.; Simpson, R.; Sorensen, C.; Zhou, M.

2017-10-01

The cost of a fusion reactor is of critical importance to its ultimate acceptability as a commercial source of electricity. While there are general rules of thumb for scaling both overnight cost and levelized cost of electricity the corresponding relations are not very accurate or universally agreed upon. We have carried out a series of scaling studies of tokamak reactor costs based on reasonably sophisticated plasma and engineering models. The analysis is largely analytic, requiring only a simple numerical code, thus allowing a very large number of designs. Importantly, the studies are aimed at plasma physicists rather than fusion engineers. The goals are to assess the pros and cons of steady state burning plasma experiments and reactors. One specific set of results discusses the benefits of higher magnetic fields, now possible because of the recent development of high T rare earth superconductors (REBCO); with this goal in mind, we calculate quantitative expressions, including both scaling and multiplicative constants, for cost and major radius as a function of central magnetic field.

Conceptual design and development of GEM based detecting system for tomographic tungsten focused transport monitoring

NASA Astrophysics Data System (ADS)

Chernyshova, M.; Czarski, T.; Malinowski, K.; Kowalska-Strzęciwilk, E.; Poźniak, K.; Kasprowicz, G.; Zabołotny, W.; Wojeński, A.; Kolasiński, P.; Mazon, D.; Malard, P.

2015-10-01

Implementing tungsten as a plasma facing material in ITER and future fusion reactors will require effective monitoring of not just its level in the plasma but also its distribution. That can be successfully achieved using detectors based on Gas Electron Multiplier (GEM) technology. This work presents the conceptual design of the detecting unit for poloidal tomography to be tested at the WEST project tokamak. The current stage of the development is discussed covering aspects which include detector's spatial dimensions, gas mixtures, window materials and arrangements inside and outside the tokamak ports, details of detector's structure itself and details of the detecting module electronics. It is expected that the detecting unit under development, when implemented, will add to the safe operation of tokamak bringing the creation of sustainable nuclear fusion reactors a step closer. A shorter version of this contribution is due to be published in PoS at: 1st EPS conference on Plasma Diagnostics

Synchronized fusion development considering physics, materials and heat transfer

NASA Astrophysics Data System (ADS)

Wong, C. P. C.; Liu, Y.; Duan, X. R.; Xu, M.; Li, Q.; Feng, K. M.; Zheng, G. Y.; Li, Z. X.; Wang, X. Y.; Li, B.; Zhang, G. S.

2017-12-01

Significant achievements have been made in the last 60 years in the development of fusion energy with the tokamak configuration. Based on the accumulated knowledge, the world is embarking on the construction and operation of ITER (International Thermonuclear Experimental Reactor) with a production of 500 MWf fusion power and the demonstration of physics Q  =  10. ITER will demonstrate D-T burn physics for a duration of a few hundred seconds to prepare for the next long-burn or steady state nuclear testing tokamak operating at much higher neutron fluence. With the evolution into a steady state nuclear device, such as the China Fusion Engineering Test Reactor (CFETR), it is necessary to examine the boundary conditions imposed by the combined development of tokamak physics, fusion materials and fusion technology for a reactor. The development of ferritic steel alloys as the structural material suitable for use at high neutron fluence leads to the use of helium as the most likely reactor coolant. This points to the fundamental technology limitation on the removal of chamber wall maximum heat flux at around 1 MW m-2 and an average heat flux of 0.1 MW m-2 for the next test reactor. Future reactor performance will then depend on the control of spatial and temporal edge heat flux peaking in order to increase the average heat flux to the chamber wall. With these severe material and technological limitations, system studies were used to scope out a few robust steady state synchronized fusion reactor (SFR) designs. As an example, a low fusion power design at 131.6 MWf, which can satisfy steady state design requirements, would have a major radius of 5.5 m and minor radius of 1.6 m. Such a design with even more advanced structural materials like W f/W composite could allow higher performance and provide a net electrical production of 62 MWe. These can be incorporated into the CFETR program.

Burn Control Mechanisms in Tokamaks

NASA Astrophysics Data System (ADS)

Hill, M. A.; Stacey, W. M.

2015-11-01

Burn control and passive safety in accident scenarios will be an important design consideration in future tokamak reactors, in particular fusion-fission hybrid reactors, e.g. the Subcritical Advanced Burner Reactor. We are developing a burning plasma dynamics code to explore various aspects of burn control, with the intent to identify feedback mechanisms that would prevent power excursions. This code solves the coupled set of global density and temperature equations, using scaling relations from experimental fits. Predictions of densities and temperatures have been benchmarked against DIII-D data. We are examining several potential feedback mechanisms to limit power excursions: i) ion-orbit loss, ii) thermal instability density limits, iii) MHD instability limits, iv) the degradation of alpha-particle confinement, v) modifications to the radial current profile, vi) ``divertor choking'' and vii) Type 1 ELMs. Work supported by the US DOE under DE-FG02-00ER54538, DE-FC02-04ER54698.

Tokamak foundation in USSR/Russia 1950-1990

NASA Astrophysics Data System (ADS)

Smirnov, V. P.

2010-01-01

In the USSR, nuclear fusion research began in 1950 with the work of I.E. Tamm, A.D. Sakharov and colleagues. They formulated the principles of magnetic confinement of high temperature plasmas, that would allow the development of a thermonuclear reactor. Following this, experimental research on plasma initiation and heating in toroidal systems began in 1951 at the Kurchatov Institute. From the very first devices with vessels made of glass, porcelain or metal with insulating inserts, work progressed to the operation of the first tokamak, T-1, in 1958. More machines followed and the first international collaboration in nuclear fusion, on the T-3 tokamak, established the tokamak as a promising option for magnetic confinement. Experiments continued and specialized machines were developed to test separately improvements to the tokamak concept needed for the production of energy. At the same time, research into plasma physics and tokamak theory was being undertaken which provides the basis for modern theoretical work. Since then, the tokamak concept has been refined by a world-wide effort and today we look forward to the successful operation of ITER.

Tungsten Deposition on Graphite using Plasma Enhanced Chemical Vapour Deposition.

NASA Astrophysics Data System (ADS)

Sharma, Uttam; Chauhan, Sachin S.; Sharma, Jayshree; Sanyasi, A. K.; Ghosh, J.; Choudhary, K. K.; Ghosh, S. K.

2016-10-01

The tokamak concept is the frontrunner for achieving controlled thermonuclear reaction on earth, an environment friendly way to solve future energy crisis. Although much progress has been made in controlling the heated fusion plasmas (temperature ∼ 150 million degrees) in tokamaks, technological issues related to plasma wall interaction topic still need focused attention. In future, reactor grade tokamak operational scenarios, the reactor wall and target plates are expected to experience a heat load of 10 MW/m2 and even more during the unfortunate events of ELM's and disruptions. Tungsten remains a suitable choice for the wall and target plates. It can withstand high temperatures, its ductile to brittle temperature is fairly low and it has low sputtering yield and low fuel retention capabilities. However, it is difficult to machine tungsten and hence usages of tungsten coated surfaces are mostly desirable. To produce tungsten coated graphite tiles for the above-mentioned purpose, a coating reactor has been designed, developed and made operational at the SVITS, Indore. Tungsten coating on graphite has been attempted and successfully carried out by using radio frequency induced plasma enhanced chemical vapour deposition (rf -PECVD) for the first time in India. Tungsten hexa-fluoride has been used as a pre-cursor gas. Energy Dispersive X-ray spectroscopy (EDS) clearly showed the presence of tungsten coating on the graphite samples. This paper presents the details of successful operation and achievement of tungsten coating in the reactor at SVITS.

Movable-molybdenum-reflector reactivity experiments for control studies of compact space power reactor concepts

NASA Technical Reports Server (NTRS)

Fox, T. A.

1973-01-01

An experimental reflector reactivity study was made with a compact cylindrical reactor using a uranyl fluoride - water fuel solution. The reactor was axially unreflected and radially reflected with segments of molybdenum. The reflector segments were displaced incrementally in both the axial and radial dimensions, and the shutdown of each configuration was measured by using the pulsed-neutron source technique. The reactivity effects for axial and radial displacement of reflector segments are tabulated separately and compared. The experiments provide data for control-system studies of compact-space-power-reactor concepts.

20 years of research on the Alcator C-Mod tokamaka)

NASA Astrophysics Data System (ADS)

Greenwald, M.; Bader, A.; Baek, S.; Bakhtiari, M.; Barnard, H.; Beck, W.; Bergerson, W.; Bespamyatnov, I.; Bonoli, P.; Brower, D.; Brunner, D.; Burke, W.; Candy, J.; Churchill, M.; Cziegler, I.; Diallo, A.; Dominguez, A.; Duval, B.; Edlund, E.; Ennever, P.; Ernst, D.; Faust, I.; Fiore, C.; Fredian, T.; Garcia, O.; Gao, C.; Goetz, J.; Golfinopoulos, T.; Granetz, R.; Grulke, O.; Hartwig, Z.; Horne, S.; Howard, N.; Hubbard, A.; Hughes, J.; Hutchinson, I.; Irby, J.; Izzo, V.; Kessel, C.; LaBombard, B.; Lau, C.; Li, C.; Lin, Y.; Lipschultz, B.; Loarte, A.; Marmar, E.; Mazurenko, A.; McCracken, G.; McDermott, R.; Meneghini, O.; Mikkelsen, D.; Mossessian, D.; Mumgaard, R.; Myra, J.; Nelson-Melby, E.; Ochoukov, R.; Olynyk, G.; Parker, R.; Pitcher, S.; Podpaly, Y.; Porkolab, M.; Reinke, M.; Rice, J.; Rowan, W.; Schmidt, A.; Scott, S.; Shiraiwa, S.; Sierchio, J.; Smick, N.; Snipes, J. A.; Snyder, P.; Sorbom, B.; Stillerman, J.; Sung, C.; Takase, Y.; Tang, V.; Terry, J.; Terry, D.; Theiler, C.; Tronchin-James, A.; Tsujii, N.; Vieira, R.; Walk, J.; Wallace, G.; White, A.; Whyte, D.; Wilson, J.; Wolfe, S.; Wright, G.; Wright, J.; Wukitch, S.; Zweben, S.

2014-11-01

The object of this review is to summarize the achievements of research on the Alcator C-Mod tokamak [Hutchinson et al., Phys. Plasmas 1, 1511 (1994) and Marmar, Fusion Sci. Technol. 51, 261 (2007)] and to place that research in the context of the quest for practical fusion energy. C-Mod is a compact, high-field tokamak, whose unique design and operating parameters have produced a wealth of new and important results since it began operation in 1993, contributing data that extends tests of critical physical models into new parameter ranges and into new regimes. Using only high-power radio frequency (RF) waves for heating and current drive with innovative launching structures, C-Mod operates routinely at reactor level power densities and achieves plasma pressures higher than any other toroidal confinement device. C-Mod spearheaded the development of the vertical-target divertor and has always operated with high-Z metal plasma facing components—approaches subsequently adopted for ITER. C-Mod has made ground-breaking discoveries in divertor physics and plasma-material interactions at reactor-like power and particle fluxes and elucidated the critical role of cross-field transport in divertor operation, edge flows and the tokamak density limit. C-Mod developed the I-mode and the Enhanced Dα H-mode regimes, which have high performance without large edge localized modes and with pedestal transport self-regulated by short-wavelength electromagnetic waves. C-Mod has carried out pioneering studies of intrinsic rotation and demonstrated that self-generated flow shear can be strong enough in some cases to significantly modify transport. C-Mod made the first quantitative link between the pedestal temperature and the H-mode's performance, showing that the observed self-similar temperature profiles were consistent with critical-gradient-length theories and followed up with quantitative tests of nonlinear gyrokinetic models. RF research highlights include direct experimental observation of ion cyclotron range of frequency (ICRF) mode-conversion, ICRF flow drive, demonstration of lower-hybrid current drive at ITER-like densities and fields and, using a set of novel diagnostics, extensive validation of advanced RF codes. Disruption studies on C-Mod provided the first observation of non-axisymmetric halo currents and non-axisymmetric radiation in mitigated disruptions. A summary of important achievements and discoveries are included.

Evaluating and planning the radioactive waste options for dismantling the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rule, K.; Scott, J.; Larson, S.

1995-12-31

The Tokamak Fusion Test Reactor (TFTR) is a one-of-a kind tritium fusion research reactor, and is planned to be decommissioned within the next several years. This is the largest fusion reactor in the world and as a result of deuterium-tritum reactions is tritium contaminated and activated from 14 Mev neutrons. This presents many unusual challenges when dismantling, packaging and disposing its components and ancillary systems. Special containers are being designed to accommodate the vacuum vessel, neutral beams, and tritium delivery and processing systems. A team of experienced professionals performed a detailed field study to evaluate the requirements and appropriate methodsmore » for packaging the radioactive materials. This team focused on several current and innovative methods for waste minimization that provides the oppurtunmost cost effective manner to package and dispose of the waste. This study also produces a functional time-phased schedule which conjoins the waste volume, weight, costs and container requirements with the detailed project activity schedule for the entire project scope. This study and project will be the first demonstration of the decommissioning of a tritium fusion test reactor. The radioactive waste disposal aspects of this project are instrumental in demonstrating the viability of a fusion power reactor with regard to its environmental impact and ultimate success.« less

Performance of compact fast pyrolysis reactor with Auger-type modules for the continuous liquid biofuel production

NASA Astrophysics Data System (ADS)

Nishimura, Shun; Ebitani, Kohki

2018-01-01

Development of a compact fast pyrolysis reactor constructed using Auger-type technology to afford liquid biofuel with high yield has been an interesting concept in support of local production for local consumption. To establish a widely useable module package, details of the performance of the developing compact module reactor were investigated. This study surveyed the properties of as-produced pyrolysis oil as a function of operation time, and clarified the recent performance of the developing compact fast pyrolysis reactor. Results show that after condensation in the scrubber collector, e.g. approx. 10 h for a 25 kg/h feedstock rate, static performance of pyrolysis oil with approximately 20 MJ/kg (4.8 kcal/g) calorific values were constantly obtained after an additional 14 h. The feeding speed of cedar chips strongly influenced the time for oil condensation process: i.e. 1.6 times higher feeding speed decreased the condensation period by half (approx. 5 h in the case of 40 kg/h). Increasing the reactor throughput capacity is an important goal for the next stage in the development of a compact fast pyrolysis reactor with Auger-type modules.

Concept of DT fuel cycle for a fusion neutron source

DOE Office of Scientific and Technical Information (OSTI.GOV)

Anan'ev, S.; Spitsyn, A.V.; Kuteev, B.V.

2015-03-15

A concept of DT-fusion neutron source (FNS) with the neutron yield higher than 10{sup 18} neutrons per second is under design in Russia. Such a FNS is of interest for many applications: 1) basic and applied research (neutron scattering, etc); 2) testing the structural materials for fusion reactors; 3) control of sub-critical nuclear systems and 4) nuclear waste processing (including transmutation of minor actinides). This paper describes the fuel cycle concept of a compact fusion neutron source based on a small spherical tokamak (FNS-ST) with a MW range of DT fusion power and considers the key physics issues of thismore » device. The major and minor radii are ∼0.5 and ∼0.3 m, magnetic field ∼1.5 T, heating power less than 15 MW and plasma current 1-2 MA. The system provides the fuel mixture with equal fractions of D and T (D:T = 1:1) for all FNS technology systems. (authors)« less

A self-consistent model of an isothermal tokamak

NASA Astrophysics Data System (ADS)

McNamara, Steven; Lilley, Matthew

2014-10-01

Continued progress in liquid lithium coating technologies have made the development of a beam driven tokamak with minimal edge recycling a feasibly possibility. Such devices are characterised by improved confinement due to their inherent stability and the suppression of thermal conduction. Particle and energy confinement become intrinsically linked and the plasma thermal energy content is governed by the injected beam. A self-consistent model of a purely beam fuelled isothermal tokamak is presented, including calculations of the density profile, bulk species temperature ratios and the fusion output. Stability considerations constrain the operating parameters and regions of stable operation are identified and their suitability to potential reactor applications discussed.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lucia, M., E-mail: mlucia@pppl.gov; Kaita, R.; Majeski, R.

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Thermal and hydraulic analysis of a cylindrical blanket module design for a tokamak reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lee, A.Y.

1978-10-01

Various existing blanket design concepts for a tokamak fusion reactor were evaluated and assessed. These included the demonstration power reactors of ORNL, GA and others. As a result of this study, a cylindrical, modularized blanket design concept was developed. The module is a double-walled, stainless steel 316 cylinder containing liquid lithium for tritium breeding and is cooled by pressurized helium. Steady state and transient thermal conditions under normal and some off-design conditions were analyzed and presented. At the steady state reference operating point the maximum structure temperature is 452/sup 0/C at the maximum stressed location and is 495/sup 0/C atmore » the less stressed location. The coolant inlet pressure is 54.4 atm, the inlet temperature is 200/sup 0/C and the exit temperature is 435/sup 0/C. The coolant could be utilized with a helium/steam turbine power conversion system with a cycle thermal efficiency of 30.8%.« less

Fusion Studies in Japan

NASA Astrophysics Data System (ADS)

Ogawa, Yuichi

2016-05-01

A new strategic energy plan decided by the Japanese Cabinet in 2014 strongly supports the steady promotion of nuclear fusion development activities, including the ITER project and the Broader Approach activities from the long-term viewpoint. Atomic Energy Commission (AEC) in Japan formulated the Third Phase Basic Program so as to promote an experimental fusion reactor project. In 2005 AEC has reviewed this Program, and discussed on selection and concentration among many projects of fusion reactor development. In addition to the promotion of ITER project, advanced tokamak research by JT-60SA, helical plasma experiment by LHD, FIREX project in laser fusion research and fusion engineering by IFMIF were highly prioritized. Although the basic concept is quite different between tokamak, helical and laser fusion researches, there exist a lot of common features such as plasma physics on 3-D magnetic geometry, high power heat load on plasma facing component and so on. Therefore, a synergetic scenario on fusion reactor development among various plasma confinement concepts would be important.

Fusion reactor blanket/shield design study

NASA Astrophysics Data System (ADS)

Smith, D. L.; Clemmer, R. G.; Harkness, S. D.; Jung, J.; Krazinski, J. L.; Mattas, R. F.; Stevens, H. C.; Youngdahl, C. K.; Trachsel, C.; Bowers, D.

1979-07-01

A joint study of Tokamak reactor first wall/blanket/shield technology was conducted to identify key technological limitations for various tritium breeding blanket design concepts, establishment of a basis for assessment and comparison of the design features of each concept, and development of optimized blanket designs. The approach used involved a review of previously proposed blanket designs, analysis of critical technological problems and design features associated with each of the blanket concepts, and a detailed evaluation of the most tractable design concepts. Tritium breeding blanket concepts were evaluated according to the proposed coolant. The effort concentrated on evaluation of lithium and water cooled blanket designs and helium and molten salt cooled designs. Generalized nuclear analysis of the tritium breeding performance, an analysis of tritium breeding requirements, and a first wall stress analysis were conducted as part of the study. The impact of coolant selection on the mechanical design of a Tokamak reactor was evaluated. Reference blanket designs utilizing the four candidate coolants are presented.

Electron Cyclotron Radiation, Related Power Loss, and Passive Current Drive in Tokamaks: A Review

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fidone, Ignazio; Giruzzi, Gerardo; Granata, Giovanni

2001-01-15

A critical review on emission of weakly damped, high-harmonics electron cyclotron radiation, the related synchrotron power loss, and passive current drive in tokamaks with a fish-scale first wall is presented. First, the properties of overlapping harmonics are discussed using general analytical formulas and numerical applications. Next, the radiation power loss and efficiency of passive current drive in tokamak reactors are derived for the asymmetric fish-scale first wall. The radiation power loss is determined by the direction-averaged reflection coefficient {sigma}{sub 0} and the passive current drive by the differential reflectivity {delta}{sigma}/(1 - {sigma}{sub 0}). Finally, the problem of experimental investigations ofmore » the high harmonics radiation spectra, of {sigma}{sub 0} and {delta}{sigma}/(1 - {sigma}{sub 0}) in existing and next-step tokamaks, is discussed. Accurate measurements of the radiation spectra and the fish-scale reflectivity can be performed at arbitrary electron temperature using a partial fish-scale structure located near the tokamak equatorial plane.« less

Impurity seeding for suppression of the near scrape-off layer heat flux feature in tokamak limited plasmas

NASA Astrophysics Data System (ADS)

Nespoli, F.; Labit, B.; Furno, I.; Theiler, C.; Sheikh, U. A.; Tsui, C. K.; Boedo, J. A.; TCV Team

2018-05-01

In inboard-limited plasmas, foreseen to be used in future fusion reactor start-up and ramp down phases, the Scrape-Off Layer (SOL) exhibits two regions: the "near" and "far" SOL. The steep radial gradient of the parallel heat flux associated with the near SOL can result in excessive thermal loads onto the solid surfaces, damaging them and/or limiting the operational space of a fusion reactor. In this article, leveraging the results presented in the study by F. Nespoli et al. [Nucl. Fusion 57, 126029 (2017)], we propose a technique for the mitigation and suppression of the near SOL heat flux feature by impurity seeding. The first successful experimental results from the TCV tokamak are presented and discussed.

On heat loading, novel divertors, and fusion reactors

NASA Astrophysics Data System (ADS)

Kotschenreuther, M.; Valanju, P. M.; Mahajan, S. M.; Wiley, J. C.

2007-07-01

The limited thermal power handling capacity of the standard divertors (used in current as well as projected tokamaks) is likely to force extremely high (˜90%) radiation fractions frad in tokamak fusion reactors that have heating powers considerably larger than ITER [D. J. Campbell, Phys. Plasmas 8, 2041 (2001)]. Such enormous values of necessary frad could have serious and debilitating consequences on the core confinement, stability, and dependability for a fusion power reactor, especially in reactors with Internal Transport Barriers. A new class of divertors, called X-divertors (XD), which considerably enhance the divertor thermal capacity through a flaring of the field lines only near the divertor plates, may be necessary and sufficient to overcome these problems and lead to a dependable fusion power reactor with acceptable economics. X-divertors will lower the bar on the necessary confinement to bring it in the range of the present experimental results. Its ability to reduce the radiative burden imparts the X-divertor with a key advantage. Lower radiation demands allow sharply peaked density profiles that enhance the bootstrap fraction creating the possibility for a highly increased beta for the same beta normal discharges. The X-divertor emerges as a beta-enhancer capable of raising it by up to roughly a factor of 2.

Characteristics of Low-q(a) Disruptions in the Compact Toroidal Hybrid

NASA Astrophysics Data System (ADS)

Pandya, M. D.; Archmiller, M. C.; Ennis, D. A.; Hartwell, G. J.; Maurer, D. A.

2014-10-01

Tokamak disruptions are dramatic events that lead to a sudden loss of plasma confinement. Disruptions that occur at low edge safety-factor, q (a) , limit the operation of tokamaks to q (a) >= 2 . The Compact Toroidal Hybrid (CTH) is a torsatron-tokamak hybrid with a helical field coil and vertical field coils to establish a stellartor equilibrium, while an ohmic coil induces plasma current. A feature of the CTH device is the ability to adjust the vacuum rotational transform, tvac (t =1/q ), by varying the ratio of current in the helical and toroidal field coils. The value of edge tvac can be varied from about 0.02 to 0.3 (qvac (a) ~ 50 to 3.3). Plasma discharges in CTH are routinely observed to operate with q (a) < 2 , and in some cases as low as q (a) ~ 1 . 1 . In CTH, low-q(a) disruptions are observed with a dominant m/n=3/2 precursor. The disruptivity of plasma discharges is over 80% when tvac (a) < 0 . 04 (qvac (a) < 25) and as tvac (a) is increased further, the disruptivity of the plasma discharges decreases. The disruptions are completely suppressed for tvac (a) > 0 . 07 (qvac (a) ~ 14) . This work is supported by US Department of Energy Grant No. DE-FG02-00ER54610.

Compact antenna for two-dimensional beam scan in the JT-60U electron cyclotron heating/current drive system

NASA Astrophysics Data System (ADS)

Moriyama, S.; Kajiwara, K.; Takahashi, K.; Kasugai, A.; Seki, M.; Ikeda, Y.; Fujii, T.

2005-11-01

A compact antenna system was designed and fabricated to enable millimeter-wave beam scanning in the toroidal and poloidal directions of the JT-60U tokamak for electron cyclotron heating (ECH) and electron cyclotron current drive (ECCD) experiments. The antenna consists of a fast movable flat mirror mounted on the tokamak vacuum vessel and a rotary focusing mirror attached at the end of the waveguide that is supported from outside the vacuum vessel. This separate support concept enables a compact structure inside a shallow port (0.68×0.54×0.2m) that is shared with a subport for an independent diagnostic system. During a plasma shot, the flat mirror is driven by a servomotor with a 3-m-long drive shaft to reduce the influence of the high magnetic field on the motor. The focusing mirror is rotated by a simple mechanism utilizing a push rod and an air cylinder. The antenna has been operated reliably for 3 years after a small improvement to the rotary drive mechanism. It has made significant contributions to ECH and ECCD experiments, especially the current profile control in JT-60U.

The radiation asymmetry in MGI rapid shutdown on J-TEXT tokamak

NASA Astrophysics Data System (ADS)

Tong, Ruihai; Chen, Zhongyong; Huang, Duwei; Cheng, Zhifeng; Zhang, Xiaolong; Zhuang, Ge; J-TEXT Team

2017-10-01

Disruptions, the sudden termination of tokamak fusion plasmas by instabilities, have the potential to cause severe material wall damage to large tokamaks like ITER. The mitigation of disruption damage is an essential part of any fusion reactor system. Massive gas injection (MGI) rapid shutdown is a technique in which large amounts of noble gas are injected into the plasma in order to safely radiate the plasma energy evenly over the entire plasma-facing first wall. However, the radiated energy during the thermal quench (TQ) in massive gas injection (MGI) induced disruptions is found toroidal asymmetric, and the degrees of asymmetry correlate with the gas penetration and MGI induced magnetohydrodynamics (MHD) activities. A toroidal and poloidal array of ultraviolet photodiodes (AXUV) has been developed to investigate the radiation asymmetry on J-TEXT tokamak. Together with the upgraded mirnov probe arrays, the relation between MGI triggered MHD activities with radiation asymmetry is studied.

Alpha-driven magnetohydrodynamics (MHD) and MHD-induced alpha loss in the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Chang, Z.; Nazikian, R.; Fu, G.Y.

1997-02-01

Alpha-driven toroidal Alfven eigenmodes (TAEs) are observed as predicted by theory in the post neutral beam phase in high central q (safety factor) deuterium-tritium (D-T) plasmas in the Tokamak Fusion Test Reactor (TFTR). The mode location, poloidal structure and the importance of q profile for TAE instability are discussed. So far no alpha particle loss due to these modes was detected due to the small mode amplitude. However, alpha loss induced by kinetic ballooning modes (KBMs) was observed in high confinement D-T discharges. Particle orbit simulation demonstrates that the wave-particle resonant interaction can explain the observed correlation between the increasemore » in alpha loss and appearance of multiple high-n (n {ge} 6, n is the toroidal mode number) modes.« less

Velocity space instabilities of alpha particles in tokamak reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sigmar, D.J.

1979-01-01

In this lecture on high frequency instability due to isotropic hollow alpha velocity distributions it was first shown that such distributions can actually arise under thermonuclear conditions in a tokamak reactor, particularly for the case of imperfect alpha particle confinement. The toroidal geometry (i.e., the poloidal variation of the alpha gyrofrequency) then leads to linear instability of the compressional Alfven wave ..omega.. = C/sub A/k/sub perpendicular/ with k/sub parallel/ congruent to O, k/sub perpendicular/ rho/sub ..cap alpha../ greater than or equal to 1, v/sub ..cap alpha../ > C/sub A/, at the low harmonics ..omega.. congruent to n ..omega../sub c..cap alpha../.more » Thus the free energy of the inverted alpha distribution is accessible and produces anomalously rapid diffusion of F/sub ..cap alpha../(v/sub perpendicular/). (MOW)« less

Tokamak power reactor ignition and time dependent fractional power operation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vold, E.L.; Mau, T.K.; Conn, R.W.

1986-06-01

A flexible time-dependent and zero-dimensional plasma burn code with radial profiles was developed and employed to study the fractional power operation and the thermal burn control options for an INTOR-sized tokamak reactor. The code includes alpha thermalization and a time-dependent transport loss which can be represented by any one of several currently popular scaling laws for energy confinement time. Ignition parameters were found to vary widely in density-temperature (n-T) space for the range of scaling laws examined. Critical ignition issues were found to include the extent of confinement time degradation by alpha heating, the ratio of ion to electron transportmore » power loss, and effect of auxiliary heating on confinement. Feedback control of the auxiliary power and ion fuel sources are shown to provide thermal stability near the ignition curve.« less

Effect of high Z impurities on the ignition and Lawson conditions for a thermonuclear reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Meade, D.M.

1973-06-01

The recent advances in plasma heating and confinement using Tokamak devices have produced plasmas which approach thermonuclear conditions. Substantial amounts (0.1 to 1%) of partially stripped high Z impurities have been observed in these discharges. These high Z impurities (Fe,Mo,W) are presumably due to plasma bombardment of the limiter and vacuum chamber wall. Since the plasma energy will be increasing sharply in the next sequence of experiments from approx. =1kJ in ST tokamak to approx. =3MJ in PLT and up to approx. =100MJ in a feasibility experiment, the bombardment of the wall and limiter will become increasingly important. In thismore » paper, the effects of high Z impurities on the ignition and Lawson conditions for a DT reactor are calculated. 7 refs., 2 figs.« less

Summary of Apollo; A D- sup 3 He tokamak reactor design

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kulcinski, G.L.; Blanchard, T.P.; El-Guebaly, L.A.

1992-07-01

In this paper, the key features of Apollo, a conceptual D-{sup 3}He tokamak reactor for commercial electricity production, are summarized. The 1000-MW (electric) design utilizes direct conversion of transport, neutron, and bremsstrahlung radiation power. The direct conversion method uses reactants, and the thermal conversion cycle uses an organic coolant. Apollo operates in the first-stability regime, with a major radius of 7.89 m, a peak magnetic field on the toroidal field coils of 19.3 T, a 53-MA plasma current, and a 6.7% beta value. The low neutron production of the D-{sup 3}He fuel cycle greatly reduces the radiation damage rate andmore » allows a full-lifetime first wall and structure made of standard steels with only slight modifications to reduce activation levels.« less

Real-time MSE measurements for current profile control on KSTAR.

PubMed

De Bock, M F M; Aussems, D; Huijgen, R; Scheffer, M; Chung, J

2012-10-01

To step up from current day fusion experiments to power producing fusion reactors, it is necessary to control long pulse, burning plasmas. Stability and confinement properties of tokamak fusion reactors are determined by the current or q profile. In order to control the q profile, it is necessary to measure it in real-time. A real-time motional Stark effect diagnostic is being developed at Korean Superconducting Tokamak for Advanced Research for this purpose. This paper focuses on 3 topics important for real-time measurements: minimize the use of ad hoc parameters, minimize external influences and a robust and fast analysis algorithm. Specifically, we have looked into extracting the retardance of the photo-elastic modulators from the signal itself, minimizing the influence of overlapping beam spectra by optimizing the optical filter design and a multi-channel, multiharmonic phase locking algorithm.

Optimization of 3D Field Design

NASA Astrophysics Data System (ADS)

Logan, Nikolas; Zhu, Caoxiang

2017-10-01

Recent progress in 3D tokamak modeling is now leveraged to create a conceptual design of new external 3D field coils for the DIII-D tokamak. Using the IPEC dominant mode as a target spectrum, the Finding Optimized Coils Using Space-curves (FOCUS) code optimizes the currents and 3D geometry of multiple coils to maximize the total set's resonant coupling. The optimized coils are individually distorted in space, creating toroidal ``arrays'' containing a variety of shapes that often wrap around a significant poloidal extent of the machine. The generalized perturbed equilibrium code (GPEC) is used to determine optimally efficient spectra for driving total, core, and edge neoclassical toroidal viscosity (NTV) torque and these too provide targets for the optimization of 3D coil designs. These conceptual designs represent a fundamentally new approach to 3D coil design for tokamaks targeting desired plasma physics phenomena. Optimized coil sets based on plasma response theory will be relevant to designs for future reactors or on any active machine. External coils, in particular, must be optimized for reliable and efficient fusion reactor designs. Work supported by the US Department of Energy under DE-AC02-09CH11466.

A current drive by using the fast wave in frequency range higher than two timeslower hybrid resonance frequency on tokamaks

NASA Astrophysics Data System (ADS)

Kim, Sun Ho; Hwang, Yong Seok; Jeong, Seung Ho; Wang, Son Jong; Kwak, Jong Gu

2017-10-01

An efficient current drive scheme in central or off-axis region is required for the steady state operation of tokamak fusion reactors. The current drive by using the fast wave in frequency range higher than two times lower hybrid resonance (w>2wlh) could be such a scheme in high density, high temperature reactor-grade tokamak plasmas. First, it has relatively higher parallel electric field to the magnetic field favorable to the current generation, compared to fast waves in other frequency range. Second, it can deeply penetrate into high density plasmas compared to the slow wave in the same frequency range. Third, parasitic coupling to the slow wave can contribute also to the current drive avoiding parametric instability, thermal mode conversion and ion heating occured in the frequency range w<2wlh. In this study, the propagation boundary, accessibility, and the energy flow of the fast wave are given via cold dispersion relation and group velocity. The power absorption and current drive efficiency are discussed qualitatively through the hot dispersion relation and the polarization. Finally, those characteristics are confirmed with ray tracing code GENRAY for the KSTAR plasmas.

Fusion pumped laser

DOEpatents

Pappas, Daniel S.

1989-01-01

Apparatus is provided for generating energy in the form of laser radiation. A tokamak fusion reactor is provided for generating a long, or continuous, pulse of high-energy neutrons. The tokamak design provides a temperature and a magnetic field which is effective to generate a neutron flux of at least 10.sup.15 neutrons/cm.sup.2.s. A conversion medium receives neutrons from the tokamak and converts the high-energy neutrons to an energy source with an intensity and an energy effective to excite a preselected lasing medium. The energy source typically comprises fission fragments, alpha particles, and radiation from a fission event. A lasing medium is provided which is responsive to the energy source to generate a population inversion which is effective to support laser oscillations for generating output radiation.

Radioactive Waste Streams: Waste Classification for Disposal

DTIC Science & Technology

2006-12-13

INL; and Fort St. Vrain, Colorado .10 In contrast to commercial reactors, naval reactors can operate without refueling for up to 20 years. 11 As of 2003...originally of the states of Arizona, Colorado , Nevada, New Mexico, Utah, and Wyoming.61 Arizona, Utah, and Wyoming later withdrew from the Compact, leaving... Colorado , Nevada, and New Mexico as remaining Compact members.62 The Rocky Mountain Compact defines low-level waste as specifically excluding

Improving spatial and spectral resolution of TCV Thomson scattering

NASA Astrophysics Data System (ADS)

Hawke, J.; Andrebe, Y.; Bertizzolo, R.; Blanchard, P.; Chavan, R.; Decker, J.; Duval, B.; Lavanchy, P.; Llobet, X.; Marlétaz, B.; Marmillod, P.; Pochon, G.; Toussaint, M.

2017-12-01

The recently completed MST2 upgrade to the Thomson scattering (TS) system on TCV (Tokamak à Configuration Variable) at the Swiss Plasma Center aims to provide an enhanced spatial and spectral resolution while maintaining the high level of diagnostic flexibility for the study of TCV plasmas. The MST2 (Medium Sized Tokamak) is a work program within the Eurofusion ITER physics department, aimed at exploiting Europe's medium sized tokamak programs for a better understanding of ITER physics. This upgrade to the TCV Thomson scattering system involved the installation of 40 new compact 5-channel spectrometers and modifications to the diagnostics fiber optic design. The complete redesign of the fiber optic backplane incorporates fewer larger diameter fibers, allowing for a higher resolution in both the core and edge of TCV plasmas along the laser line, with a slight decrease in the signal to noise ratio of Thomson measurements. The 40 new spectrometers added to the system are designed to cover the full range of temperatures expected in TCV, able to measure electron temperatures (Te) with high precision between (6 eV and 20 keV) . The design of these compact spectrometers stems originally from the design utilized in the MAST (Mega Amp Spherical Tokamak) TS system located in Oxfordshire, United Kingdom. This design was implemented on TCV with an overall layout of optical fibers and spectrometers to achieve an overall increase in the spatial resolution, specifically a resolution of approximately 1% of the minor radius within the plasma pedestal region. These spectrometers also enhance the diagnostic spectral resolution, especially within the plasma edge, due to the low Te measurement capabilities. These additional spectrometers allow for a much greater diagnostic flexibility, allowing for quality full Thomson profiles in 75% of TCV plasma configurations.

Compact power reactor

DOEpatents

Wetch, Joseph R.; Dieckamp, Herman M.; Wilson, Lewis A.

1978-01-01

There is disclosed a small compact nuclear reactor operating in the epithermal neutron energy range for supplying power at remote locations, as for a satellite. The core contains fuel moderator elements of Zr hydride with 7 w/o of 93% enriched uranium alloy. The core has a radial beryllium reflector and is cooled by liquid metal coolant such as NaK. The reactor is controlled and shut down by moving portions of the reflector.

METHOD OF PREPARING A FUEL ELEMENT FOR A NUCLEAR REACTOR

DOEpatents

Hauth, J.J.; Anicetti, R.J.

1962-12-01

A method is described for preparing a fuel element for a nuclear reactor. According to the patent uranium dioxide is compacted in a metal tabe by directlng intense sound waves at the tabe prior to tamp packing or vibration compaction of the powder. (AEC)

Parameter exploration for a Compact Advanced Tokamak DEMO

NASA Astrophysics Data System (ADS)

Weisberg, D. B.; Buttery, R. J.; Ferron, J. R.; Garofalo, A. M.; Snyder, P. B.; Turnbull, A. D.; Holcomb, C. T.; McClenaghan, J.; Canik, J.; Park, J.-M.

2017-10-01

A new parameter study has explored a range of design points to assess the physics feasibility for a compact 200MWe advanced tokamak DEMO that combines high beta (βN < 4) and high toroidal field (BT = 6 - 7 T). A unique aspect of this study is the use of a FASTRAN modeling suite that combines integrated transport, pedestal, stability, and heating & current drive calculations to predict steady-state solutions with neutral beam and helicon powered current drive. This study has identified a range of design solutions in a compact (R0 = 4 m), high-field (BT = 6 - 7 T), strongly-shaped (κ = 2 , δ = 0.6) device. Unlike previous proposals, C-AT DEMO takes advantage of high-beta operation as well as emerging advances in magnet technology to demonstrate net electric production in a moderately sized machine. We present results showing that the large bootstrap fraction and low recirculating power enabled by high normalized beta can achieve tolerable heat and neutron load with good H-mode access. The prediction of operating points with simultaneously achieved high-confinement (H98 < 1.3), high-density (fGW < 1.3), and high-beta warrants additional assessment of this approach towards a cost-attractive DEMO device. Work supported by the US DOE under DE-FC02-04ER54698.

Control of impurities in toroidal plasma devices

DOEpatents

Ohkawa, Tihiro

1980-01-01

A method and apparatus for plasma impurity control in closed flux plasma systems such as Tokamak reactors is disclosed. Local axisymmetrical injection of hydrogen gas is employed to reverse the normally inward flow of impurities into the plasma.

Preliminary consideration of CFETR ITER-like case diagnostic system.

PubMed

Li, G S; Yang, Y; Wang, Y M; Ming, T F; Han, X; Liu, S C; Wang, E H; Liu, Y K; Yang, W J; Li, G Q; Hu, Q S; Gao, X

2016-11-01

Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basic control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.

Preliminary consideration of CFETR ITER-like case diagnostic system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, G. S.; Liu, Y. K.; Gao, X.

2016-11-15

Chinese Fusion Engineering Test Reactor (CFETR) is a new superconducting tokamak device being designed in China, which aims at bridging the gap between ITER and DEMO, where DEMO is a tokamak demonstration fusion reactor. Two diagnostic cases, ITER-like case and towards DEMO case, have been considered for CFETR early and later operating phases, respectively. In this paper, some preliminary consideration of ITER-like case will be presented. Based on ITER diagnostic system, three versions of increased complexity and coverage of the ITER-like case diagnostic system have been developed with different goals and functions. Version A aims only machine protection and basicmore » control. Both of version B and version C are mainly for machine protection, basic and advanced control, but version C has an increased level of redundancy necessary for improved measurements capability. The performance of these versions and needed R&D work are outlined.« less

Current drive at plasma densities required for thermonuclear reactors.

PubMed

Cesario, R; Amicucci, L; Cardinali, A; Castaldo, C; Marinucci, M; Panaccione, L; Santini, F; Tudisco, O; Apicella, M L; Calabrò, G; Cianfarani, C; Frigione, D; Galli, A; Mazzitelli, G; Mazzotta, C; Pericoli, V; Schettini, G; Tuccillo, A A

2010-08-10

Progress in thermonuclear fusion energy research based on deuterium plasmas magnetically confined in toroidal tokamak devices requires the development of efficient current drive methods. Previous experiments have shown that plasma current can be driven effectively by externally launched radio frequency power coupled to lower hybrid plasma waves. However, at the high plasma densities required for fusion power plants, the coupled radio frequency power does not penetrate into the plasma core, possibly because of strong wave interactions with the plasma edge. Here we show experiments performed on FTU (Frascati Tokamak Upgrade) based on theoretical predictions that nonlinear interactions diminish when the peripheral plasma electron temperature is high, allowing significant wave penetration at high density. The results show that the coupled radio frequency power can penetrate into high-density plasmas due to weaker plasma edge effects, thus extending the effective range of lower hybrid current drive towards the domain relevant for fusion reactors.

TFTR CAMAC systems and components

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rauch, W.A.; Bergin, W.; Sichta, P.

1987-08-01

Princeton's tokamak fusion test reactor (TFTR) utilizes Computer Automated Measurement and Control (CAMAC) to provide instrumentation for real and quasi real time control, monitoring, and data acquisition systems. This paper describes and discusses the complement of CAMAC hardware systems and components that comprise the interface for tokamak control and measurement instrumentation, and communication with the central instrumentation control and data acquisition (CICADA) system. It also discusses CAMAC reliability and calibration, types of modules used, a summary of data acquisition and control points, and various diagnostic maintenance tools used to support and troubleshoot typical CAMAC systems on TFTR.

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

NASA Astrophysics Data System (ADS)

Adamek, J.; Müller, H. W.; Silva, C.; Schrittwieser, R.; Ionita, C.; Mehlmann, F.; Costea, S.; Horacek, J.; Kurzan, B.; Bilkova, P.; Böhm, P.; Aftanas, M.; Vondracek, P.; Stöckel, J.; Panek, R.; Fernandes, H.; Figueiredo, H.

2016-04-01

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (ΦBPP) and the floating potential (Vfl) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula Te = (ΦBPP - Vfl)/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes.

PubMed

Adamek, J; Müller, H W; Silva, C; Schrittwieser, R; Ionita, C; Mehlmann, F; Costea, S; Horacek, J; Kurzan, B; Bilkova, P; Böhm, P; Aftanas, M; Vondracek, P; Stöckel, J; Panek, R; Fernandes, H; Figueiredo, H

2016-04-01

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (ΦBPP) and the floating potential (Vfl) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula Te = (ΦBPP - Vfl)/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.

The computation in diagnostics for tokamaks: systems, designs, approaches

NASA Astrophysics Data System (ADS)

Krawczyk, Rafał; Linczuk, Paweł; Czarski, Tomasz; Wojeński, Andrzej; Chernyshova, Maryna; Poźniak, Krzysztof; Kolasiński, Piotr; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol; Gaska, Michał

2017-08-01

The requirements given for GEM (Gaseous Electron Multiplier) detector based acquisition system for plasma impurities diagnostics triggered a need for the development of a specialized software and hardware architecture. The amount of computations with latency and throughput restrictions cause that an advanced solution is sought for. In order to provide a mechanism fitting the designated tokamaks, an insight into existing solutions was necessary. In the article there is discussed architecture of systems used for plasma diagnostics and in related scientific fields. The developed solution is compared and contrasted with other diagnostic and control systems. Particular attention is payed to specific requirements for plasma impurities diagnostics in tokamak thermal fusion reactor. Subsequently, the details are presented that justified the choice of the system architecture and the discussion on various approaches is given.

Energetic electrons, hard x-ray emission and MHD activity studies in the IR-T1 tokamak.

PubMed

Agah, K Mikaili; Ghoranneviss, M; Elahi, A Salar

2015-01-01

Determinations of plasma parameters as well as the Magnetohydrodynamics (MHD) activity, energetic electrons energy and energy confinement time are essential for future fusion reactors experiments and optimized operation. Also some of the plasma information can be deduced from these parameters, such as plasma equilibrium, stability, and MHD instabilities. In this contribution we investigated the relation between energetic electrons, hard x-ray emission and MHD activity in the IR-T1 Tokamak. For this purpose we used the magnetic diagnostics and a hard x-ray spectroscopy in IR-T1 tokamak. A hard x-ray emission is produced by collision of the runaway electrons with the plasma particles or limiters. The mean energy was calculated from the slope of the energy spectrum of hard x-ray photons.

Development progress of the Materials Analysis and Particle Probe

NASA Astrophysics Data System (ADS)

Lucia, M.; Kaita, R.; Majeski, R.; Bedoya, F.; Allain, J. P.; Boyle, D. P.; Schmitt, J. C.; Onge, D. A. St.

2014-11-01

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Development progress of the Materials Analysis and Particle Probe.

PubMed

Lucia, M; Kaita, R; Majeski, R; Bedoya, F; Allain, J P; Boyle, D P; Schmitt, J C; Onge, D A St

2014-11-01

The Materials Analysis and Particle Probe (MAPP) is a compact in vacuo surface science diagnostic, designed to provide in situ surface characterization of plasma facing components in a tokamak environment. MAPP has been implemented for operation on the Lithium Tokamak Experiment at Princeton Plasma Physics Laboratory (PPPL), where all control and analysis systems are currently under development for full remote operation. Control systems include vacuum management, instrument power, and translational/rotational probe drive. Analysis systems include onboard Langmuir probes and all components required for x-ray photoelectron spectroscopy, low-energy ion scattering spectroscopy, direct recoil spectroscopy, and thermal desorption spectroscopy surface analysis techniques.

Li/sub 2/O microsphere fabrication. Monthly letter progress report No. 1 for the period ending May 31, 1977

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schluderberg, D C

1977-06-01

The purpose of this task is to establish process parameters for the fabrication of lithium oxide (Li/sub 2/O) microspheres having properties which as closely as possible approximate those required for the design characteristics of the University of Wisconsin design for a TOKAMAK-type fusion reactor utilizing the moving bed of Li/sub 2/O microspheres as both reactor coolant and tritium breeder.

Theoretical and experimental studies of a planar inductive coupled rf plasma source as the driver in simulator facility (ISTAPHM) of interactions of waves with the edge plasma on tokamaks

NASA Astrophysics Data System (ADS)

Ghanei, V.; Nasrabadi, M. N.; Chin, O.-H.; Jayapalan, K. K.

2017-11-01

This research aims to design and build a planar inductive coupled RF plasma source device which is the driver of the simulator project (ISTAPHM) of the interactions between ICRF Antenna and Plasma on tokamak by using the AMPICP model. For this purpose, a theoretical derivation of the distribution of the RF magnetic field in the plasma-filled reactor chamber is presented. An experimental investigation of the field distributions is described and Langmuir measurements are developed numerically. A comparison of theory and experiment provides an evaluation of plasma parameters in the planar ICP reactor. The objective of this study is to characterize the plasma produced by the source alone. We present the results of the first analysis of the plasma characteristics (plasma density, electron temperature, electron-ion collision frequency, particle fluxes and their velocities, stochastic frequency, skin depth and electron energy distribution functions) as function of the operating parameters (injected power, neutral pressure and magnetic field) as measured with fixed and movable Langmuir probes. The plasma is currently produced only by the planar ICP. The exact goal of these experiments is that the produced plasma by external source can exist as a plasma representative of the edge of tokamaks.

Current Challenges in the First Principle Quantitative Modelling of the Lower Hybrid Current Drive in Tokamaks

NASA Astrophysics Data System (ADS)

Peysson, Y.; Bonoli, P. T.; Chen, J.; Garofalo, A.; Hillairet, J.; Li, M.; Qian, J.; Shiraiwa, S.; Decker, J.; Ding, B. J.; Ekedahl, A.; Goniche, M.; Zhai, X.

2017-10-01

The Lower Hybrid (LH) wave is widely used in existing tokamaks for tailoring current density profile or extending pulse duration to steady-state regimes. Its high efficiency makes it particularly attractive for a fusion reactor, leading to consider it for this purpose in ITER tokamak. Nevertheless, if basics of the LH wave in tokamak plasma are well known, quantitative modeling of experimental observations based on first principles remains a highly challenging exercise, despite considerable numerical efforts achieved so far. In this context, a rigorous methodology must be carried out in the simulations to identify the minimum number of physical mechanisms that must be considered to reproduce experimental shot to shot observations and also scalings (density, power spectrum). Based on recent simulations carried out for EAST, Alcator C-Mod and Tore Supra tokamaks, the state of the art in LH modeling is reviewed. The capability of fast electron bremsstrahlung, internal inductance li and LH driven current at zero loop voltage to constrain all together LH simulations is discussed, as well as the needs of further improvements (diagnostics, codes, LH model), for robust interpretative and predictive simulations.

Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

DOE PAGES

Bartel, N.; Chen, M.; Utgikar, V. P.; ...

2015-04-04

A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

Comparative analysis of compact heat exchangers for application as the intermediate heat exchanger for advanced nuclear reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bartel, N.; Chen, M.; Utgikar, V. P.

A comparative evaluation of alternative compact heat exchanger designs for use as the intermediate heat exchanger in advanced nuclear reactor systems is presented in this article. Candidate heat exchangers investigated included the Printed circuit heat exchanger (PCHE) and offset strip-fin heat exchanger (OSFHE). Both these heat exchangers offer high surface area to volume ratio (a measure of compactness [m2/m3]), high thermal effectiveness, and overall low pressure drop. Helium–helium heat exchanger designs for different heat exchanger types were developed for a 600 MW thermal advanced nuclear reactor. The wavy channel PCHE with a 15° pitch angle was found to offer optimummore » combination of heat transfer coefficient, compactness and pressure drop as compared to other alternatives. The principles of the comparative analysis presented here will be useful for heat exchanger evaluations in other applications as well.« less

Resistive edge mode instability in stellarator and tokamak geometries

NASA Astrophysics Data System (ADS)

Mahmood, M. Ansar; Rafiq, T.; Persson, M.; Weiland, J.

2008-09-01

Geometrical effects on linear stability of electrostatic resistive edge modes are investigated in the three-dimensional Wendelstein 7-X stellarator [G. Grieger et al., Plasma Physics and Controlled Nuclear Fusion Research 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 3, p. 525] and the International Thermonuclear Experimental Reactor [Progress in the ITER Physics Basis, Nucl. Fusion 7, S1, S285 (2007)]-like equilibria. An advanced fluid model is used for the ions together with the reduced Braghinskii equations for the electrons. Using the ballooning mode representation, the drift wave problem is set as an eigenvalue equation along a field line and is solved numerically using a standard shooting technique. A significantly larger magnetic shear and a less unfavorable normal curvature in the tokamak equilibrium are found to give a stronger finite-Larmor radius stabilization and a more narrow mode spectrum than in the stellarator. The effect of negative global magnetic shear in the tokamak is found to be stabilizing. The growth rate on a tokamak magnetic flux surface is found to be comparable to that on a stellarator surface with the same global magnetic shear but the eigenfunction in the tokamak is broader than in the stellarator due to the presence of large negative local magnetic shear (LMS) on the tokamak surface. A large absolute value of the LMS in a region of unfavorable normal curvature is found to be stabilizing in the stellarator, while in the tokamak case, negative LMS is found to be stabilizing and positive LMS destabilizing.

Catalytic reactor

DOEpatents

Aaron, Timothy Mark [East Amherst, NY; Shah, Minish Mahendra [East Amherst, NY; Jibb, Richard John [Amherst, NY

2009-03-10

A catalytic reactor is provided with one or more reaction zones each formed of set(s) of reaction tubes containing a catalyst to promote chemical reaction within a feed stream. The reaction tubes are of helical configuration and are arranged in a substantially coaxial relationship to form a coil-like structure. Heat exchangers and steam generators can be formed by similar tube arrangements. In such manner, the reaction zone(s) and hence, the reactor is compact and the pressure drop through components is minimized. The resultant compact form has improved heat transfer characteristics and is far easier to thermally insulate than prior art compact reactor designs. Various chemical reactions are contemplated within such coil-like structures such that as steam methane reforming followed by water-gas shift. The coil-like structures can be housed within annular chambers of a cylindrical housing that also provide flow paths for various heat exchange fluids to heat and cool components.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bourham, Mohamed A.; Gilligan, John G.

Safety considerations in large future fusion reactors like ITER are important before licensing the reactor. Several scenarios are considered hazardous, which include safety of plasma-facing components during hard disruptions, high heat fluxes and thermal stresses during normal operation, accidental energy release, and aerosol formation and transport. Disruption events, in large tokamaks like ITER, are expected to produce local heat fluxes on plasma-facing components, which may exceed 100 GW/m{sup 2} over a period of about 0.1 ms. As a result, the surface temperature dramatically increases, which results in surface melting and vaporization, and produces thermal stresses and surface erosion. Plasma-facing componentsmore » safety issues extends to cover a wide range of possible scenarios, including disruption severity and the impact of plasma-facing components on disruption parameters, accidental energy release and short/long term LOCA's, and formation of airborne particles by convective current transport during a LOVA (water/air ingress disruption) accident scenario. Study, and evaluation of, disruption-induced aerosol generation and mobilization is essential to characterize database on particulate formation and distribution for large future fusion tokamak reactor like ITER. In order to provide database relevant to ITER, the SIRENS electrothermal plasma facility at NCSU has been modified to closely simulate heat fluxes expected in ITER.« less

Laser or charged-particle-beam fusion reactor with direct electric generation by magnetic flux compression

DOEpatents

Lasche, G.P.

1983-09-29

The invention is a laser or particle-beam-driven fusion reactor system which takes maximum advantage of both the very short pulsed nature of the energy release of inertial confinement fusion (ICF) and the very small volumes within which the thermonuclear burn takes place. The pulsed nature of ICF permits dynamic direct energy conversion schemes such as magnetohydrodynamic (MHD) generation and magnetic flux compression; the small volumes permit very compact blanket geometries. By fully exploiting these characteristics of ICF, it is possible to design a fusion reactor with exceptionally high power density, high net electric efficiency, and low neutron-induced radioactivity. The invention includes a compact blanket design and method and apparatus for obtaining energy utilizing the compact blanket.

Experiments with lithium limiter on T-11M tokamak and applications of the lithium capillary-pore system in future fusion reactor devices

NASA Astrophysics Data System (ADS)

Mirnov, S. V.; Azizov, E. A.; Evtikhin, V. A.; Lazarev, V. B.; Lyublinski, I. E.; Vertkov, A. V.; Prokhorov, D. Yu

2006-06-01

The paper is an overview of recent results of Li limiter testing in T-11M tokamak. The lithium limiter is based on the capillary-pore system (CPS) concept. The Li erosion process and deuterium (D2) and helium (He) sorption by Li first wall were investigated. The ability of capillary forces to confine the liquid Li in the CPS limiter during disruption was demonstrated. The idea of combined lithium limiter with thin (0.6 mm) CPS coating as a solution of the heat removal problem was realized. As a result the quasi steady-state tokamak regime with duration up to 0.3 s and clean (Zeff = 1) deuterium plasma has been achieved. The temporal evolution of the lithium surface temperature during discharge was measured by a IR radiometer and then was recalculated to the surface power load. For the estimation of the Li limiter erosion the Li neutral and ions spectral line emission were observed. The increase in lithium erosion as a result of limiter heating was discovered. The radial distribution of plasma column radiation measurements showed up to 90% of the total radiation losses in a relatively thin (5 cm) boundary layer and only 10% in a plasma centre during discharges with high Li influx. Oscillations of Li emission and saw-tooth-like oscillations of the limiter surface temperature have been detected in discharge regimes with highest Li limiter temperature (>600 °C). A version of Li CPS first wall of DEMO reactor and Li CPS limiter experiment in the International Thermonuclear Energy Reactor are suggested.

Profile measurements of the electron temperature on the ASDEX Upgrade, COMPASS, and ISTTOK tokamak using Thomson scattering, triple, and ball-pen probes

DOE Office of Scientific and Technical Information (OSTI.GOV)

Adamek, J., E-mail: adamek@ipp.cas.cz; Horacek, J.; Bilkova, P.

The ball-pen probe (BPP) technique is used successfully to make profile measurements of the electron temperature on the ASDEX Upgrade (Axially Symmetric Divertor Experiment), COMPASS (COMPact ASSembly), and ISTTOK (Instituto Superior Tecnico TOKamak) tokamak. The electron temperature is provided by a combination of the BPP potential (Φ{sub BPP}) and the floating potential (V{sub fl}) of the Langmuir probe (LP), which is compared with the Thomson scattering diagnostic on ASDEX Upgrade and COMPASS. Excellent agreement between the two diagnostics is obtained for circular and diverted plasmas and different heating mechanisms (Ohmic, NBI, ECRH) in deuterium discharges with the same formula T{submore » e} = (Φ{sub BPP} − V{sub fl})/2.2. The comparative measurements of the electron temperature using BPP/LP and triple probe (TP) techniques on the ISTTOK tokamak show good agreement of averaged values only inside the separatrix. It was also found that the TP provides the electron temperature with significantly higher standard deviation than BPP/LP. However, the resulting values of both techniques are well in the phase with the maximum of cross-correlation function being 0.8.« less

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

DOE PAGES

Okabayashi, M.; Zanca, P.; Strait, E. J.; ...

2016-11-25

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Okabayashi, M.; Zanca, P.; Strait, E. J.

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. We have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance ismore » provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. We define β N as β N = β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. Finally, the internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.« less

Avoidance of tearing mode locking with electro-magnetic torque introduced by feedback-based mode rotation control in DIII-D and RFX-mod

NASA Astrophysics Data System (ADS)

Okabayashi, M.; Zanca, P.; Strait, E. J.; Garofalo, A. M.; Hanson, J. M.; In, Y.; La Haye, R. J.; Marrelli, L.; Martin, P.; Paccagnella, R.; Paz-Soldan, C.; Piovesan, P.; Piron, C.; Piron, L.; Shiraki, D.; Volpe, F. A.; DIII-D, The; RFX-mod Teams

2017-01-01

Disruptions caused by tearing modes (TMs) are considered to be one of the most critical roadblocks to achieving reliable, steady-state operation of tokamak fusion reactors. Here we have demonstrated a promising scheme to avoid mode locking by utilizing the electro-magnetic (EM) torque produced with 3D coils that are available in many tokamaks. In this scheme, the EM torque is delivered to the modes by a toroidal phase shift between the externally applied field and the excited TM fields, compensating for the mode momentum loss through the interaction with the resistive wall and uncorrected error fields. Fine control of torque balance is provided by a feedback scheme. We have explored this approach in two widely different devices and plasma conditions: DIII-D and RFX-mod operated in tokamak mode. In DIII-D, the plasma target was high β N in a non-circular divertor tokamak. Here β N is defined as β N  =  β/(I p /aB t) (%Tm/MA), where β, I p, a, B t are the total stored plasma pressure normalized by the magnetic pressure, plasma current, plasma minor radius and toroidal magnetic field at the plasma center, respectively. The RFX-mod plasma was ohmically-heated with ultra-low safety factor in a circular limiter discharge with active feedback coils outside the thick resistive shell. The DIII-D and RFX-mod experiments showed remarkable consistency with theoretical predictions of torque balance. The application to ignition-oriented devices such as the International Thermonuclear Experimental Reactor (ITER) would expand the horizon of its operational regime. The internal 3D coil set currently under consideration for edge localized mode suppression in ITER would be well suited for this purpose.

Neutral helium beam probe

NASA Astrophysics Data System (ADS)

Karim, Rezwanul

1999-10-01

This article discusses the development of a code where diagnostic neutral helium beam can be used as a probe. The code solves numerically the evolution of the population densities of helium atoms at their several different energy levels as the beam propagates through the plasma. The collisional radiative model has been utilized in this numerical calculation. The spatial dependence of the metastable states of neutral helium atom, as obtained in this numerical analysis, offers a possible diagnostic tool for tokamak plasma. The spatial evolution for several hypothetical plasma conditions was tested. Simulation routines were also run with the plasma parameters (density and temperature profiles) similar to a shot in the Princeton beta experiment modified (PBX-M) tokamak and a shot in Tokamak Fusion Test Reactor tokamak. A comparison between the simulation result and the experimentally obtained data (for each of these two shots) is presented. A good correlation in such comparisons for a number of such shots can establish the accurateness and usefulness of this probe. The result can possibly be extended for other plasma machines and for various plasma conditions in those machines.

Progress toward steady-state tokamak operation exploiting the high bootstrap current fraction regime

DOE PAGES

Ren, Q. L.; Garofalo, A. M.; Gong, X. Z.; ...

2016-06-20

Recent DIII-D experiments have increased the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Improved understanding of scenario stability has led to the achievement of very high values of β p and β N despite strong ITBs. Good confinement has been achieved with reduced toroidal rotation. These high β p plasmas challenge the energy transport understanding, especiallymore » in the electron energy channel. A new turbulent transport model, named 2 TGLF-SAT1, has been developed which improves the transport prediction. Experiments extending results to long pulse on EAST, based on the physics basis developed at DIII-D, have been conducted. Finally, more investigations will be carried out on EAST with more additional auxiliary power to come online in the near term.« less

Tungsten coating by ATC plasma spraying on CFC for WEST tokamak

NASA Astrophysics Data System (ADS)

Firdaouss, M.; Desgranges, C.; Hernandez, C.; Mateus, C.; Maier, H.; Böswirth, B.; Greuner, H.; Samaille, F.; Bucalossi, J.; Missirlian, M.

2017-12-01

In the field of fusion experiments using a tokamak, the plasma facing components (PFC) are the closest object to the hot plasma. Due to the plasma-wall interaction, the material composing the PFC may enter the plasma and disturb the experiments. In the past, the main material for PFC was carbon (CFC, graphite), while the future reactors like ITER will be fully metallic, in particular tungsten. The Tore Supra tokamak has been transformed in an x-point divertor fusion device within the frame of the WEST (W (tungsten) Environment in Steady-state Tokamak) project in order to have plasma conditions close to those expected in ITER. The PFC other than the divertor has been coated with W to transform Tore Supra into a fully metallic environment. Different coating techniques have been selected for different kind of PFC. This paper gives an overview on the coating process used for the antennae protection limiter, the associated validation programme and concludes on the adequacy of the W coating with the WEST experimental programme requirements and gives perspectives on the development to be pursued.

Control of magnetohydrodynamic stability by phase space engineering of energetic ions in tokamak plasmas.

PubMed

Graves, J P; Chapman, I T; Coda, S; Lennholm, M; Albergante, M; Jucker, M

2012-01-10

Virtually collisionless magnetic mirror-trapped energetic ion populations often partially stabilize internally driven magnetohydrodynamic disturbances in the magnetosphere and in toroidal laboratory plasma devices such as the tokamak. This results in less frequent but dangerously enlarged plasma reorganization. Unique to the toroidal magnetic configuration are confined 'circulating' energetic particles that are not mirror trapped. Here we show that a newly discovered effect from hybrid kinetic-magnetohydrodynamic theory has been exploited in sophisticated phase space engineering techniques for controlling stability in the tokamak. These theoretical predictions have been confirmed, and the technique successfully applied in the Joint European Torus. Manipulation of auxiliary ion heating systems can create an asymmetry in the distribution of energetic circulating ions in the velocity orientated along magnetic field lines. We show the first experiments in which large sawtooth collapses have been controlled by this technique, and neoclassical tearing modes avoided, in high-performance reactor-relevant plasmas.

The physics of tokamak start-up

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mueller, D.

Tokamak start-up on present-day devices usually relies on inductively induced voltage from a central solenoid. In some cases, inductive startup is assisted with auxiliary power from electron cyclotron radio frequency heating. International Thermonuclear Experimental Reactor, the National Spherical Torus Experiment Upgrade and JT60, now under construction, will make use of the understanding gained from present-day devices to ensure successful start-up. Design of a spherical tokamak (ST) with DT capability for nuclear component testing would require an alternative to a central solenoid because the small central column in an ST has insufficient space to provide shielding for the insulators in themore » solenoid. Alternative start-up techniques such as induction using outer poloidal field coils, electron Bernstein wave start-up, coaxial helicity injection, and point source helicity injection have been used with success, but require demonstration of scaling to higher plasma current.« less

Numerical verification of bounce-harmonic resonances in neoclassical toroidal viscosity for tokamaks.

PubMed

Kim, Kimin; Park, Jong-Kyu; Boozer, Allen H

2013-05-03

This Letter presents the first numerical verification for the bounce-harmonic (BH) resonance phenomena of the neoclassical transport in a tokamak perturbed by nonaxisymmetric magnetic fields. The BH resonances were predicted by analytic theories of neoclassical toroidal viscosity (NTV), as the parallel and perpendicular drift motions can be resonant and result in a great enhancement of the radial momentum transport. A new drift-kinetic δf guiding-center particle code, POCA, clearly verified that the perpendicular drift motions can reduce the transport by phase-mixing, but in the BH resonances the motions can form closed orbits and particles radially drift out fast. The POCA calculations on resulting NTV torque are largely consistent with analytic calculations, and show that the BH resonances can easily dominate the NTV torque when a plasma rotates in the perturbed tokamak and therefore, is a critical physics for predicting the rotation and stability in the International Thermonuclear Experimental Reactor.

Reactor application of an improved bundle divertor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Yang, T.F.; Ruck, G.W.; Lee, A.Y.

1978-11-01

A Bundle Divertor was chosen as the impurity control and plasma exhaust system for the beam driven Demonstration Tokamak Hybrid Reactor - DTHR. In the context of a preconceptual design study of the reactor and associated facility a bundle divertor concept was developed and integrated into the reactor system. The overall system was found feasible and scalable for reactors with intermediate torodial field strengths on axis. The important design characteristics are: the overall average current density of the divertor coils is 0.73 kA for each tesla of toroidal field on axis; the divertor windings are made from super-conducting cables supportedmore » by steel structures and are designed to be maintainable; the particle collection assembly and auxiliary cryosorption vacuum pump are dual systems designed such that they can be reactivated alterntively to allow for continuous reactor operation; and the power requirement for energizing and operating the divertor is about 5 MW.« less

Alternative approaches to plasma confinement

NASA Technical Reports Server (NTRS)

Roth, J. R.

1978-01-01

The paper discusses 20 plasma confinement schemes each representing an alternative to the tokamak fusion reactor. Attention is given to: (1) tokamak-like devices (TORMAC, Topolotron, and the Extrap concept), (2) stellarator-like devices (Torsatron and twisted-coil stellarators), (3) mirror machines (Astron and reversed-field devices, the 2XII B experiment, laser-heated solenoids, the LITE experiment, the Kaktus-Surmac concept), (4) bumpy tori (hot electron bumpy torus, toroidal minimum-B configurations), (5) electrostatically assisted confinement (electrostatically stuffed cusps and mirrors, electrostatically assisted toroidal confinement), (6) the Migma concept, and (7) wall-confined plasmas. The plasma parameters of the devices are presented and the advantages and disadvantages of each are listed.

Spectrum lines of highly ionized zinc, germanium, selenium, zirconium, molybdenum, and silver injected into Princeton Large Torus and Tokamak Fusion Test Reactor tokamak discharges

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hinnov, E.; Boody, F.; Cohen, S.

1986-10-01

Measured wavelengths of a number of highly ionized atoms are reported. These include the 3s/sup 2/3p--3s3p/sup 2/ and 3s/sup 2/3p--3s/sup 2/3d transitions in the aluminum isoelectronic sequence of Zn XVIII, Ge XX, Se XXII, Zr XXVIII, Mo XXX, and Ag XXXV; several transitions in the n = 2 shell of Zn XXII, Zn XXIII, and Zn XXIV; and the resonance and intercombination lines of Ag XXXVI--Ag XXXVII and of Ge XXIX--Ge XXX.

The Compact Ignition Tokamak and electron cyclotron heating: Description of need; assessment of prospects

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ignat, D.W.; Cohn, D.R.; Woskov, P.P.

1989-01-01

The CIT will benefit from auxiliary heating of 10 to 40 MW. The schedules of both the CIT construction project and the operating plan contain adequate time to develop and implement ECH systems based on the gyrotron and the induction free electron laser (IFEL). Each approach has advantages and is the object of R and D at the level of many millions of dollars per year. While the gyrotron is further advanced in terms of power and pulse length achieved, rapid progress is scheduled for the IFEL, including experiments on tokamaks. Plans of CIT, gyrotron, and IFEL make 1992 anmore » appropriate time frame to commit to one or both systems. 12 refs., 8 figs., 2 tabs.« less

Energetic ion loss detector on the Alcator C-Mod tokamak.

PubMed

Pace, D C; Granetz, R S; Vieira, R; Bader, A; Bosco, J; Darrow, D S; Fiore, C; Irby, J; Parker, R R; Parkin, W; Reinke, M L; Terry, J L; Wolfe, S M; Wukitch, S J; Zweben, S J

2012-07-01

A scintillator-based energetic ion loss detector has been successfully commissioned on the Alcator C-Mod tokamak. This probe is located just below the outer midplane, where it captures ions of energies up to 2 MeV resulting from ion cyclotron resonance heating. After passing through a collimating aperture, ions impact different regions of the scintillator according to their gyroradius (energy) and pitch angle. The probe geometry and installation location are determined based on modeling of expected lost ions. The resulting probe is compact and resembles a standard plasma facing tile. Four separate fiber optic cables view different regions of the scintillator to provide phase space resolution. Evolving loss levels are measured during ion cyclotron resonance heating, including variation dependent upon individual antennae.

Deuterium-tritium experiments on the Tokamak Fusion Test reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hosea, J.; Adler, J.H.; Alling, P.

The deuterium-tritium (D-T) experimental program on the Tokamak Fusion Test Reactor (TFTR) is underway and routine tritium operations have been established. The technology upgrades made to the TFTR facility have been demonstrated to be sufficient for supporting both operations and maintenance for an extended D-T campaign. To date fusion power has been increased to {approx}9 MW and several physics results of importance to the D-T reactor regime have been obtained: electron temperature, ion temperature, and plasma stored energy all increase substantially in the D-T regime relative to the D-D regime at the same neutral beam power and comparable limiter conditioning;more » possible alpha electron heating is indicated and energy confinement improvement with average ion mass is observed; and alpha particle losses appear to be classical with no evidence of TAE mode activity up to the PFUS {approx}6 MW level. Instability in the TAE mode frequency range has been observed at PFUS > 7 MW and its effect on performance in under investigation. Preparations are underway to enhance the alpha particle density further by increasing fusion power and by extending the neutral beam pulse length to permit alpha particle effects of relevance to the ITER regime to be more fully explored.« less

Core plasma design of the compact helical reactor with a consideration of the equipartition effect

NASA Astrophysics Data System (ADS)

Goto, T.; Miyazawa, J.; Yanagi, N.; Tamura, H.; Tanaka, T.; Sakamoto, R.; Suzuki, C.; Seki, R.; Satake, S.; Nunami, M.; Yokoyama, M.; Sagara, A.; the FFHR Design Group

2018-07-01

Integrated physics analysis of plasma operation scenario of the compact helical reactor FFHR-c1 has been conducted. The DPE method, which predicts radial profiles in a reactor by direct extrapolation from the reference experimental data, has been extended to implement the equipartition effect. Close investigation of the plasma operation regime has been conducted and a candidate plasma operation point of FFHR-c1 has been identified within the parameter regime that has already been confirmed in LHD experiment in view of MHD equilibrium, MHD stability and neoclassical transport.

Compact Torus Acceleration and Injection Experiment

NASA Astrophysics Data System (ADS)

Fukumoto, Naoyuki; Fujiwara, Makoto; Nagata, Masayoshi; Uyama, Tadao; Oda, Yasushi; Azuma, Kingo

1996-11-01

The spheromak-type compact torus (CT) acceleration and injection experiment has been carried out using the Himeji Institute of Technology Compact Torus Injector (HIT-CTI). We explore the possibility of refueling, density control, current drive, and edge electric field control of tokamak plasma by means of CT injection. In last September the new HIT-CTI was built up to achieve higher speed (Vct>200 km/s) and higher density CT plasmoid by improving the capacitor bank system and eliminating the impurity and neutral particles. At initial formation discharge tests the gun for formation and compression successfully produced a CT plasmoid and injected it between electrodes for acceleration. (Initial velocity Vct.ini. 32 km/s, Bct 1 kG, Rct=5.5 cm). The formation capacitor bank will be upgraded to two 36 mF capacitors operating at 20 kV (14.4 kJ). The acceleration capacitor bank will be also upgraded to two 36 mF capacitors operating at 20 kV (14.4 kJ). The HIT-CTI will be optimized to obtain suitable CT parameters after acceleration (Bct>5 kG, Lct 20 cm, Vct>200 km/s). In the respect of CT parameter measurement magnetic probes and a He-Ne laser interferometer will be employed in order to measure the CT magnetic field, velocity, density, and length. CT acceleration experimental data on the HIT-CTI and the plan of CT injection experiment on the JFT-2M tokamak (JAERI) will be presented at the meeting.

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Brooks; A.H. Reiman; G.H. Neilson

High-beta, low-aspect-ratio (compact) stellarators are promising solutions to the problem of developing a magnetic plasma configuration for magnetic fusion power plants that can be sustained in steady-state without disrupting. These concepts combine features of stellarators and advanced tokamaks and have aspect ratios similar to those of tokamaks (2-4). They are based on computed plasma configurations that are shaped in three dimensions to provide desired stability and transport properties. Experiments are planned as part of a program to develop this concept. A beta = 4% quasi-axisymmetric plasma configuration has been evaluated for the National Compact Stellarator Experiment (NCSX). It has amore » substantial bootstrap current and is shaped to stabilize ballooning, external kink, vertical, and neoclassical tearing modes without feedback or close-fitting conductors. Quasi-omnigeneous plasma configurations stable to ballooning modes at beta = 4% have been evaluated for the Quasi-Omnigeneous Stellarator (QOS) experiment. These equilibria have relatively low bootstrap currents and are insensitive to changes in beta. Coil configurations have been calculated that reconstruct these plasma configurations, preserving their important physics properties. Theory- and experiment-based confinement analyses are used to evaluate the technical capabilities needed to reach target plasma conditions. The physics basis for these complementary experiments is described.« less

Safety and Environment aspects of Tokamak- type Fusion Power Reactor- An Overview

NASA Astrophysics Data System (ADS)

Doshi, Bharat; Reddy, D. Chenna

2017-04-01

Naturally occurring thermonuclear fusion reaction (of light atoms to form a heavier nucleus) in the sun and every star in the universe, releases incredible amounts of energy. Demonstrating the controlled and sustained reaction of deuterium-tritium plasma should enable the development of fusion as an energy source here on Earth. The promising fusion power reactors could be operated on the deuterium-tritium fuel cycle with fuel self-sufficiency. The potential impact of fusion power on the environment and the possible risks associated with operating large-scale fusion power plants is being studied by different countries. The results show that fusion can be a very safe and sustainable energy source. A fusion power plant possesses not only intrinsic advantages with respect to safety compared to other sources of energy, but also a negligible long term impact on the environment provided certain precautions are taken in its design. One of the important considerations is in the selection of low activation structural materials for reactor vessel. Selection of the materials for first wall and breeding blanket components is also important from safety issues. It is possible to fully benefit from the advantages of fusion energy if safety and environmental concerns are taken into account when considering the conceptual studies of a reactor design. The significant safety hazards are due to the tritium inventory and energetic neutron fluence induced activity in the reactor vessel, first wall components, blanket system etc. The potential of release of radioactivity under operational and accident conditions needs attention while designing the fusion reactor. Appropriate safety analysis for the quantification of the risk shall be done following different methods such as FFMEA (Functional Failure Modes and Effects Analysis) and HAZOP (Hazards and operability). Level of safety and safety classification such as nuclear safety and non-nuclear safety is very important for the FPR (Fusion Power Reactor). This paper describes an overview of safety and environmental merits of fusion power reactor, issues and design considerations and need for R&D on safety and environmental aspects of Tokamak type fusion reactor.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lazarus, E; Peng, Yueng Kay Martin

Oak Ridge National Laboratory (ORNL) proposes to build the Spherical Torus Experiment (STX), a very low aspect ratio toroidal confinement device. This proposal concentrates on tokamak operation of the experiment; however, it can in principle be operated as a pinch or reversed-field pinch as well. As a tokamak, the spherical torus confines a plasma that is characterized by high toroidal beta, low poloidal beta, large natural elongation, high plasma current for a given edge q, and strong paramagnetism. These features combine to offer the possibility of a compact, low-field fusion device. The figure below shows that when compared to amore » conventional tokamak the spherical torus represents a major change in geometry. The primary goals of the experiment will be to demonstrate a capability for high beta (20%) in the first stability regime, to extend our knowledge of tokamak confinement scaling, and to test oscillating-field current drive. The experiment will operate in the high-beta, collisionless regime, which is achieved in STX at low temperatures because of the geometry. At a minimum, operation of STX will help to resolve fundamental questions regarding the scaling of beta and confinement in tokamaks. Complete success in this program would have a significant impact on toroidal fusion research in that it would demonstrate solutions to the problems of beta and steady-state operation in the tokamak. The proposed device has a major radius of 0.45 m, a toroidai field of 0.5 T, a plasma current of 900 kA, and heating by neutral beam injection. We estimate 30 months for design, construction, and assembly. The budget estimate, including contingency and escalation, is $6.8 million.« less

Fuel cycle for a fusion neutron source

NASA Astrophysics Data System (ADS)

Ananyev, S. S.; Spitsyn, A. V.; Kuteev, B. V.

2015-12-01

The concept of a tokamak-based stationary fusion neutron source (FNS) for scientific research (neutron diffraction, etc.), tests of structural materials for future fusion reactors, nuclear waste transmutation, fission reactor fuel production, and control of subcritical nuclear systems (fusion-fission hybrid reactor) is being developed in Russia. The fuel cycle system is one of the most important systems of FNS that provides circulation and reprocessing of the deuterium-tritium fuel mixture in all fusion reactor systems: the vacuum chamber, neutral injection system, cryogenic pumps, tritium purification system, separation system, storage system, and tritium-breeding blanket. The existing technologies need to be significantly upgraded since the engineering solutions adopted in the ITER project can be only partially used in the FNS (considering the capacity factor higher than 0.3, tritium flow up to 200 m3Pa/s, and temperature of reactor elements up to 650°C). The deuterium-tritium fuel cycle of the stationary FNS is considered. The TC-FNS computer code developed for estimating the tritium distribution in the systems of FNS is described. The code calculates tritium flows and inventory in tokamak systems (vacuum chamber, cryogenic pumps, neutral injection system, fuel mixture purification system, isotope separation system, tritium storage system) and takes into account tritium loss in the fuel cycle due to thermonuclear burnup and β decay. For the two facility versions considered, FNS-ST and DEMO-FNS, the amount of fuel mixture needed for uninterrupted operation of all fuel cycle systems is 0.9 and 1.4 kg, consequently, and the tritium consumption is 0.3 and 1.8 kg per year, including 35 and 55 g/yr, respectively, due to tritium decay.

Current-drive by lower hybrid waves in the presence of energetic alpha-particles

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fisch, N.J.; Rax, J.M.

1991-10-01

Many experiments have now proved the effectiveness of lower hybrid waves for driving toroidal current in tokamaks. The use of these waves, however, to provide all the current in a reactor is thought to be uncertain because the waves may not penetrate the center of the more energetic reactor plasma, and, if they did, the wave power may be absorbed by alpha particles rather than by electrons. This paper explores the conditions under which lower-hybrid waves might actually drive all the current. 26 refs.

Spheromak reactor-design study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Les, J.M.

1981-06-30

A general overview of spheromak reactor characteristics, such as MHD stability, start up, and plasma geometry is presented. In addition, comparisons are made between spheromaks, tokamaks and field reversed mirrors. The computer code Sphero is also discussed. Sphero is a zero dimensional time independent transport code that uses particle confinement times and profile parameters as input since they are not known with certainty at the present time. More specifically, Sphero numerically solves a given set of transport equations whose solutions include such variables as fuel ion (deuterium and tritium) density, electron density, alpha particle density and ion, electron temperatures.

Optimization study of normal conductor tokamak for commercial neutron source

NASA Astrophysics Data System (ADS)

Fujita, T.; Sakai, R.; Okamoto, A.

2017-05-01

The optimum conceptual design of tokamak with normal conductor coils was studied for minimizing the cost for producing a given neutron flux by using a system code, PEC. It is assumed that the fusion neutrons are used for burning transuranics from the fission reactor spent fuel in the blanket and a fraction of the generated electric power is circulated to opearate the tokamak with moderate plasma fusion gain. The plasma performance was assumed to be moderate ones; {β\\text{N}}~∼ ~3{--}4 in the aspect ratio A~=~2{--}3 and {{H}98y2}~=~1 . The circulating power is an important factor affecting the cost. Though decreasing the aspect ratio is useful to raise the plasma beta and decrease the toroidal field, the maximum field in the coil starts to rise in the very low aspect ratio range and then the circulating power increases with decrease in the plasma aspect ratio A below A~∼ ~2 , while the construction cost increases with A . As a result, the cost per neutron has its minimum around A~∼ ~2.2 , namely, between ST and the conventional tokamak. The average circulating power fraction is expected to be ~51%.

Observation of finite-. beta. MHD phenomena in tokamaks

DOE Office of Scientific and Technical Information (OSTI.GOV)

McGuire, K.M.

1984-09-01

Stable high-beta plasmas are required for the tokamak to attain an economical fusion reactor. Recently, intense neutral beam heating experiments in tokamaks have shown new effects on plasma stability and confinement associated with high beta plasmas. The observed spectrum of MHD fluctuations at high beta is clearly dominated by the n = 1 mode when the q = 1 surface is in the plasma. The m/n = 1/1 mode drives other n = 1 modes through toroidal coupling and n > 1 modes through nonlinear coupling. On PDX, with near perpendicular injection, a resonant interaction between the n = 1more » internal kink and the trapped fast ions results in loss of beam particles and heating power. Key parameters in the theory are the value of q/sub 0/ and the injection angle. High frequency broadband magnetic fluctuations have been observed on ISX-B and D-III and a correlation with the deterioration of plasma confinement was reported. During enhanced confinement (H-mode) discharges in divertor plasmas, two new edge instabilities were observed, both localized radially near the separatrix. By assembling results from the different tokamak experiments, it is found that the simple theoretical ideal MHD beta limit has not been exceeded. Whether this represents an ultimate tokamak limit or if beta optimized configurations (Dee- or bean-shaped plasmas) can exceed this limit and perhaps enter a second regime of stability remains to be clarified.« less

Dynamic optimization of open-loop input signals for ramp-up current profiles in tokamak plasmas

NASA Astrophysics Data System (ADS)

Ren, Zhigang; Xu, Chao; Lin, Qun; Loxton, Ryan; Teo, Kok Lay

2016-03-01

Establishing a good current spatial profile in tokamak fusion reactors is crucial to effective steady-state operation. The evolution of the current spatial profile is related to the evolution of the poloidal magnetic flux, which can be modeled in the normalized cylindrical coordinates using a parabolic partial differential equation (PDE) called the magnetic diffusion equation. In this paper, we consider the dynamic optimization problem of attaining the best possible current spatial profile during the ramp-up phase of the tokamak. We first use the Galerkin method to obtain a finite-dimensional ordinary differential equation (ODE) model based on the original magnetic diffusion PDE. Then, we combine the control parameterization method with a novel time-scaling transformation to obtain an approximate optimal parameter selection problem, which can be solved using gradient-based optimization techniques such as sequential quadratic programming (SQP). This control parameterization approach involves approximating the tokamak input signals by piecewise-linear functions whose slopes and break-points are decision variables to be optimized. We show that the gradient of the objective function with respect to the decision variables can be computed by solving an auxiliary dynamic system governing the state sensitivity matrix. Finally, we conclude the paper with simulation results for an example problem based on experimental data from the DIII-D tokamak in San Diego, California.

Beyond ITER: neutral beams for a demonstration fusion reactor (DEMO) (invited).

PubMed

McAdams, R

2014-02-01

In the development of magnetically confined fusion as an economically sustainable power source, International Tokamak Experimental Reactor (ITER) is currently under construction. Beyond ITER is the demonstration fusion reactor (DEMO) programme in which the physics and engineering aspects of a future fusion power plant will be demonstrated. DEMO will produce net electrical power. The DEMO programme will be outlined and the role of neutral beams for heating and current drive will be described. In particular, the importance of the efficiency of neutral beam systems in terms of injected neutral beam power compared to wallplug power will be discussed. Options for improving this efficiency including advanced neutralisers and energy recovery are discussed.

ORNL-TNS/PEPR overall heating requirements

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peng, Y. K.M.; Rome, J. A.

1977-01-01

The ORNL TNS/PEPR studies have the objectives of (1) leading to a system that demonstrates the fusion reactor core in the mid-to-late 1980's and extrapolates to an economic tokamak power reactor, and (2) providing a near-term focus for the scientific and technological programs toward the power reactor. This discussion of the overall heating requirements for the ORNL TNS/PEPR is concerned with the neutral beams as the primary heating method, the electron-cyclotron resonance (ECR) heating at a lower power level for profile control, and the upper hybrid resonance (UHR) initiation and preheating of currentless plasmas to reduce current start-up loop voltagemore » (V/sub l/) requirements.« less

The engineering design of the Tokamak Physics Experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Schmidt, J.A.

A mission and supporting physics objectives have been developed, which establishes an important role for the Tokamak Physics Experiment (TPX) in developing the physic basis for a future fusion reactor. The design of TPX include advanced physics features, such as shaping and profile control, along with the capability of operating for very long pulses. The development of the superconducting magnets, actively cooled internal hardware, and remote maintenance will be an important technology contribution to future fusion projects, such as ITER. The Conceptual Design and Management Systems for TPX have been developed and reviewed, and the project is beginning Preliminary Design.more » If adequately funded the construction project should be completed in the year 2000.« less

Effect of heating on the suppression of tearing modes in tokamaks.

PubMed

Classen, I G J; Westerhof, E; Domier, C W; Donné, A J H; Jaspers, R J E; Luhmann, N C; Park, H K; van de Pol, M J; Spakman, G W; Jakubowski, M W

2007-01-19

The suppression of (neoclassical) tearing modes is of great importance for the success of future fusion reactors like ITER. Electron cyclotron waves can suppress islands, both by driving noninductive current in the island region and by heating the island, causing a perturbation to the Ohmic plasma current. This Letter reports on experiments on the TEXTOR tokamak, investigating the effect of heating, which is usually neglected. The unique set of tools available on TEXTOR, notably the dynamic ergodic divertor to create islands with a fully known driving term, and the electron cyclotron emission imaging diagnostic to provide detailed 2D electron temperature information, enables a detailed study of the suppression process and a comparison with theory.

Microinstability properties of negative magnetic shear discharges in the Tokamak Fusion Test Reactor and DIII-D

DOE Office of Scientific and Technical Information (OSTI.GOV)

Rewoldt, G.; Tang, W.M.; Lao, L.L.

1997-03-01

The microinstability properties of discharges with negative (reversed) magnetic shear in the Tokamak Fusion Test Reactor (TFTR) and DIII-D experiments with and without confinement transitions are investigated. A comprehensive kinetic linear eigenmode calculation employing the ballooning representation is employed with experimentally measured profile data, and using the corresponding numerically computed magnetohydrodynamic (MHD) equilibria. The instability considered is the toroidal drift mode (trapped-electron-{eta}{sub i} mode). A variety of physical effects associated with differing q-profiles are explained. In addition, different negative magnetic shear discharges at different times in the discharge for TFTR and DIII-D are analyzed. The effects of sheared toroidal rotation,more » using data from direct spectroscopic measurements for carbon, are analyzed using comparisons with results from a two-dimensional calculation. Comparisons are also made for nonlinear stabilization associated with shear in E{sub r}/RB{sub {theta}}. The relative importance of changes in different profiles (density, temperature, q, rotation, etc.) on the linear growth rates is considered.« less

Fusion Safety Program annual report, fiscal year 1994

NASA Astrophysics Data System (ADS)

Longhurst, Glen R.; Cadwallader, Lee C.; Dolan, Thomas J.; Herring, J. Stephen; McCarthy, Kathryn A.; Merrill, Brad J.; Motloch, Chester C.; Petti, David A.

1995-03-01

This report summarizes the major activities of the Fusion Safety Program in fiscal year 1994. The Idaho National Engineering Laboratory (INEL) is the designated lead laboratory and Lockheed Idaho Technologies Company is the prime contractor for this program. The Fusion Safety Program was initiated in 1979. Activities are conducted at the INEL, at other DOE laboratories, and at other institutions, including the University of Wisconsin. The technical areas covered in this report include tritium safety, beryllium safety, chemical reactions and activation product release, safety aspects of fusion magnet systems, plasma disruptions, risk assessment failure rate data base development, and thermalhydraulics code development and their application to fusion safety issues. Much of this work has been done in support of the International Thermonuclear Experimental Reactor (ITER). Also included in the report are summaries of the safety and environmental studies performed by the Fusion Safety Program for the Tokamak Physics Experiment and the Tokamak Fusion Test Reactor and of the technical support for commercial fusion facility conceptual design studies. A major activity this year has been work to develop a DOE Technical Standard for the safety of fusion test facilities.

Development of tf coil support concepts by design methodology in the case of a Bitter-type magnet. [Bitter-type magnets

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brossmann, U.B.

1981-01-01

The application of the methodological design is demonstrated for the development of support concepts in the case of a Bitter-type magnet designed for a compact tokamak experimentat aiming at ignition of a DT plasma. With this methodology all boundary conditions and design criteria are more easily satisfied in a technical and economical way.

Ultrafast two-dimensional lithium beam emission spectroscopy diagnostic on the EAST tokamak

NASA Astrophysics Data System (ADS)

Zoletnik, S.; Hu, G. H.; Tál, B.; Dunai, D.; Anda, G.; Asztalos, O.; Pokol, G. I.; Kálvin, S.; Németh, J.; Krizsanóczi, T.

2018-06-01

A diagnostic instrument is described for the Experimental Advanced Superconducting Tokamak (EAST) for the measurement of the edge plasma electron density profile and plasma turbulence properties. An accelerated neutral lithium beam is injected into the tokamak and the Doppler shifted 670.8 nm light emission of the Li2p-2s transition is detected. A novel compact setup is used, where the beam injection and observation take place from the same equatorial diagnostic port and radial-poloidal resolution is achieved with microsecond time resolution. The observation direction is optimized in order to achieve a sufficient Doppler shift of the beam light to be able to separate from the strong edge lithium line emission on this lithium coated device. A 250 kHz beam chopping technique is also demonstrated for the removal of background light. First results show the capability of measuring turbulence and its poloidal flow velocity in the scrape-off layer and edge region and the resolution of details of transient phenomena like edge localized modes with few microsecond time resolution.

Fishbone activity in experimental advanced superconducting tokamak neutral beam injection plasma

DOE Office of Scientific and Technical Information (OSTI.GOV)

Xu, Liqing; Zhang, Jizong; Chen, Kaiyun, E-mail: Kychen@ipp.cas.cn, E-mail: lqhu@ipp.cas.cn

2015-12-15

Repetitive fishbones near the trapped ion procession frequency were observed for the first time in the neutral beam injection high confinement plasmas in Experimental Advanced Superconducting Tokamak (EAST) tokamak, and diagnosed using a solid-state neutral particle analyzer based on a compact silicon photodiode together with an upgraded high spatial-temporal-resolution multi-arrays soft X-ray (SX) system. This 1/1 typical internal kink mode propagates in the ion-diamagnetism direction with a rotation speed faster than the bulk plasma in the plasma frame. From the SX measurements, this mode frequency is typical of chirping down and the energetic particle effect related to the twisting modemore » structure. This ion fishbone was found able to trigger a multiple core sawtooth crashes with edge-2/1 sideband modes, as well as to lead to a transition from fishbone to long lived saturated kink mode to fishbone. Furthermore, using SX tomography, a correlation between mode amplitude and mode frequency was found. Finally, a phenomenological prey–predator model was found to reproduce the fishbone nonlinear process well.« less

The High Field Path to Practical Fusion Energy

NASA Astrophysics Data System (ADS)

Mumgaard, Robert; Whyte, D.; Greenwald, M.; Hartwig, Z.; Brunner, D.; Sorbom, B.; Marmar, E.; Minervini, J.; Bonoli, P.; Irby, J.; Labombard, B.; Terry, J.; Vieira, R.; Wukitch, S.

2017-10-01

We propose a faster, lower cost development path for fusion energy enabled by high temperature superconductors, devices at high magnetic field, innovative technologies and modern approaches to technology development. Timeliness, scale, and economic-viability are the drivers for fusion energy to combat climate change and aid economic development. The opportunities provided by high-temperature superconductors, innovative engineering and physics, and new organizational structures identified over the last few years open new possibilities for realizing practical fusion energy that could meet mid-century de-carbonization needs. We discuss re-factoring the fusion energy development path with an emphasis on concrete risk retirement strategies utilizing a modular approach based on the high-field tokamak that leverages the broader tokamak physics understanding of confinement, stability, and operational limits. Elements of this plan include development of high-temperature superconductor magnets, simplified immersion blankets, advanced long-leg divertors, a compact divertor test tokamak, efficient current drive, modular construction, and demountable magnet joints. An R&D plan culminating in the construction of an integrated pilot plant and test facility modeled on the ARC concept is presented.

Proceedings of a Symposium on Advanced Compact Reactor Systems

NASA Technical Reports Server (NTRS)

1983-01-01

Reactor system technologies suitable for a variety of aerospace and terrestrial applications are considered. Technologies, safety and regulatory considerations, potential applications, and research and development opportunities are covered.

Skyshine study for next generation of fusion devices

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gohar, Y.; Yang, S.

1987-02-01

A shielding analysis for next generation of fusion devices (ETR/INTOR) was performed to study the dose equivalent outside the reactor building during operation including the contribution from neutrons and photons scattered back by collisions with air nuclei (skyshine component). Two different three-dimensional geometrical models for a tokamak fusion reactor based on INTOR design parameters were developed for this study. In the first geometrical model, the reactor geometry and the spatial distribution of the deuterium-tritium neutron source were simplified for a parametric survey. The second geometrical model employed an explicit representation of the toroidal geometry of the reactor chamber and themore » spatial distribution of the neutron source. The MCNP general Monte Carlo code for neutron and photon transport was used to perform all the calculations. The energy distribution of the neutron source was used explicitly in the calculations with ENDF/B-V data. The dose equivalent results were analyzed as a function of the concrete roof thickness of the reactor building and the location outside the reactor building.« less

3D toroidal physics: Testing the boundaries of symmetry breakinga)

NASA Astrophysics Data System (ADS)

Spong, Donald A.

2015-05-01

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to provide the plasma control needed for a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D edge localized mode suppression fields to stellarators with more dominant 3D field structures. This motivates the development of physics models that are applicable across the full range of 3D devices. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with the requirements of future fusion reactors.

Numerical Solution of the Electron Heat Transport Equation and Physics-Constrained Modeling of the Thermal Conductivity via Sequential Quadratic Programming Optimization in Nuclear Fusion Plasmas

NASA Astrophysics Data System (ADS)

Paloma, Cynthia S.

The plasma electron temperature (Te) plays a critical role in a tokamak nu- clear fusion reactor since temperatures on the order of 108K are required to achieve fusion conditions. Many plasma properties in a tokamak nuclear fusion reactor are modeled by partial differential equations (PDE's) because they depend not only on time but also on space. In particular, the dynamics of the electron temperature is governed by a PDE referred to as the Electron Heat Transport Equation (EHTE). In this work, a numerical method is developed to solve the EHTE based on a custom finite-difference technique. The solution of the EHTE is compared to temperature profiles obtained by using TRANSP, a sophisticated plasma transport code, for specific discharges from the DIII-D tokamak, located at the DIII-D National Fusion Facility in San Diego, CA. The thermal conductivity (also called thermal diffusivity) of the electrons (Xe) is a plasma parameter that plays a critical role in the EHTE since it indicates how the electron temperature diffusion varies across the minor effective radius of the tokamak. TRANSP approximates Xe through a curve-fitting technique to match experimentally measured electron temperature profiles. While complex physics-based model have been proposed for Xe, there is a lack of a simple mathematical model for the thermal diffusivity that could be used for control design. In this work, a model for Xe is proposed based on a scaling law involving key plasma variables such as the electron temperature (Te), the electron density (ne), and the safety factor (q). An optimization algorithm is developed based on the Sequential Quadratic Programming (SQP) technique to optimize the scaling factors appearing in the proposed model so that the predicted electron temperature and magnetic flux profiles match predefined target profiles in the best possible way. A simulation study summarizing the outcomes of the optimization procedure is presented to illustrate the potential of the proposed modeling method.

Physics of the Tokamak Pedestal, and Implications for Magnetic Fusion Energy

NASA Astrophysics Data System (ADS)

Snyder, Philip

2017-10-01

High performance in tokamaks is achieved via the spontaneous formation of a transport barrier in the outer few percent of the confined plasma. This narrow insulating layer, referred to as a ``pedestal,'' typically results in a >30x increase in pressure across a 0.4-5cm layer. Predicted fusion power scales with the square of the pedestal top pressure (or ``pedestal height''), hence a fusion reactor strongly benefits from a high pedestal, provided this can be attained without large Edge Localized Modes (ELMs), which may erode plasma facing materials. The overlap of drift orbit, turbulence, and equilibrium scales across this narrow layer leads to rich and complex physics, and challenges traditional analytic and computational approaches. We review studies employing gyrokinetic, neoclassical, MHD, and other methods, which have explored how a range of instabilities, influenced by complex geometry, and strong ExB flows and bootstrap current, drive transport across the pedestal and guide its structure and dynamics. Development of high resolution diagnostics, and coordinated experiments on several tokamaks, have validated understanding of important aspects of the physics, while highlighting open issues. A predictive model (EPED) has proven capable of predicting the pedestal height and width to 20-25% accuracy in large statistical studies. This model was used to predict a new, high pedestal ``Super H-Mode'' regime, which was subsequently discovered on DIII-D, and motivated experiments on Alcator C-Mod which achieved world record, reactor relevant pedestal pressure. We review open issues including improved formalism, particle and momentum transport, the role of neutrals and impurities, ELM control, and pedestal formation. Finally we discuss coupling pedestal and core predictive models to enable more comprehensive optimization of the tokamak fusion concept. Supported by the US DOE under DE-FG02-95ER54309, FC02-06ER54873, DE-FC02-04ER54698, DE-FC02-99ER54512.

MHD Effects of a Ferritic Wall on Tokamak Plasmas

NASA Astrophysics Data System (ADS)

Hughes, Paul E.

It has been recognized for some time that the very high fluence of fast (14.1MeV) neutrons produced by deuterium-tritium fusion will represent a major materials challenge for the development of next-generation fusion energy projects such as a fusion component test facility and demonstration fusion power reactor. The best-understood and most promising solutions presently available are a family of low-activation steels originally developed for use in fission reactors, but the ferromagnetic properties of these steels represent a danger to plasma confinement through enhancement of magnetohydrodynamic instabilities and increased susceptibility to error fields. At present, experimental research into the effects of ferromagnetic materials on MHD stability in toroidal geometry has been confined to demonstrating that it is still possible to operate an advanced tokamak in the presence of ferromagnetic components. In order to better quantify the effects of ferromagnetic materials on tokamak plasma stability, a new ferritic wall has been installated in the High Beta Tokamak---Extended Pulse (HBT-EP) device. The development, assembly, installation, and testing of this wall as a modular upgrade is described, and the effect of the wall on machine performance is characterized. Comparative studies of plasma dynamics with the ferritic wall close-fitting against similar plasmas with the ferritic wall retracted demonstrate substantial effects on plasma stability. Resonant magnetic perturbations (RMPs) are applied, demonstrating a 50% increase in n = 1 plasma response amplitude when the ferritic wall is near the plasma. Susceptibility of plasmas to disruption events increases by a factor of 2 or more with the ferritic wall inserted, as disruptions are observed earlier with greater frequency. Growth rates of external kink instabilities are observed to be twice as large in the presence of a close-fitting ferritic wall. Initial studies are made of the influence of mode rotation frequency on the ferritic effect, as well as observations of the effect of the ferritic wall on disruption halo currents.

What`s fair is fair

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nachtrieb, R.; Freidberg, J.P.

The newly elucidated strategy for the magnetic fusion program set forth by the Department of Energy calls for increased emphasis on alternate concepts. This strategy is motivated by the recognition that in spite of its many attractive features, a tokamak tends to be a low power density device, ultimately translating into large and corresponding expensive reactor. ITER, as it is currently envisaged, is a good example of a large, expensive, plain vanilla tokamak. In its defense, ITER rightly claims that its base design is very conservative in order to minimize the risk of failure. In order to increase power densitymore » and reduce cost there are two qualitatively different approaches that one can follow: discover advanced modes of tokamak operation or develop near alternate concepts. To decide which path to follow is a difficult task because of the uncertainties involved in making accurate comparisons between different concepts at different stages of development. One area, however, that most would agree is meaningful is ideal MHD stability. For any given concept to be credible as a reactor, it must at least be stable against macroscopic ideal MHD modes. The TPX design, for instance, goes to considerable trouble to obtain stability against external kinks: a close fitting metallic cage, rotation to stabilize the resistive wall version of the external kink, and, if all else fails, feedback. For credibility any other advanced tokamak or alternate concept should be held to the same standards of ideal MHD stability. As a first step in addressing this requirement we have investigated the stability of the RFP since it can be simply and accurately modeled as a straight cylinder. The RFP is well known to have good stability at high P against internal modes but is very unstable to external modes. We have developed a linear stability code which treats the plasma as an ideal compressible fluid, and includes longitudinal flow and a resistive wall.« less

Compact "diode-based" multi-energy soft x-ray diagnostic for NSTX.

PubMed

Tritz, K; Clayton, D J; Stutman, D; Finkenthal, M

2012-10-01

A novel and compact, diode-based, multi-energy soft x-ray (ME-SXR) diagnostic has been developed for the National Spherical Tokamak Experiment. The new edge ME-SXR system tested on NSTX consists of a set of vertically stacked diode arrays, each viewing the plasma tangentially through independent pinholes and filters providing an overlapping view of the plasma midplane which allows simultaneous SXR measurements with coarse sub-sampling of the x-ray spectrum. Using computed x-ray spectral emission data, combinations of filters can provide fast (>10 kHz) measurements of changes in the electron temperature and density profiles providing a method to "fill-in" the gaps of the multi-point Thomson scattering system.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Peng, Y.K.M.; Strickler, D.J.

The spherical torus is a very small aspect ratio (A < 2) confinement concept obtained by retaining only the indispensable components inboard to the plasma torus. MHD equilibrium calculations show that spherical torus plasmas with safety factor q > 2 are characterized by high toroidal beta (..beta../sub t/ > 0.2), low poloidal beta (..beta../sub p/ < 0.3), naturally large elongation (kappa greater than or equal to 2), large plasma current with I/sub p//(aB/sub t0/) up to about 7 MA/mT, strong paramagnetism (B/sub t//B/sub t0/ > 1.5), and strong plasma helicity (F comparable to THETA). A large near-omnigeneous region is seenmore » at the large-major-radius, bad-curvature region of the plasma in comparison with the conventional tokamaks. These features combine to engender the spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost. Because of its strong paramagnetism and helicity, the spherical torus plasma shares some of the desirable features of spheromak and reversed-field pinch (RFP) plasmas, but with tokamak-like confinement and safety factor q. The general class of spherical tori, which includes the spherical tokamak (q > 1), the spherical pinch (1 > q > O), and the spherical RFP (q < O), have magnetic field configurations unique in comparison with conventional tokamaks and RFPs. 22 refs., 12 figs.« less

Neoclassical Theory and Its Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Shaing, Ker-Chung

2015-11-20

The grant entitled Neoclassical Theory and Its Applications started on January 15 2001 and ended on April 14 2015. The main goal of the project is to develop neoclassical theory to understand tokamak physics, and employ it to model current experimental observations and future thermonuclear fusion reactors. The PI had published more than 50 papers in refereed journals during the funding period.

Tokamak experimental power reactor conceptual design. Volume II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1976-08-01

Volume II contains the following appendices: (1) summary of EPR design parameters, (2) impurity control, (3) plasma computational models, (4) structural support system, (5) materials considerations for the primary energy conversion system, (6) magnetics, (7) neutronics penetration analysis, (8) first wall stress analysis, (9) enrichment of isotopes of hydrogen by cryogenic distillation, and (10) noncircular plasma considerations. (MOW)

Simulating the effects of stellarator geometry on gyrokinetic drift-wave turbulence

NASA Astrophysics Data System (ADS)

Baumgaertel, Jessica Ann

Nuclear fusion is a clean, safe form of energy with abundant fuel. In magnetic fusion energy (MFE) experiments, the plasma fuel is confined by magnetic fields at very high temperatures and densities. One fusion reactor design is the non-axisymmetric, torus-shaped stellarator. Its fully-3D fields have advantages over the simpler, better-understood axisymmetric tokamak, including the ability to optimize magnetic configurations for desired properties, such as lower transport (longer confinement time). Turbulence in the plasma can break MFE confinement. While turbulent transport is known to cause a significant amount of heat loss in tokamaks, it is a new area of research in stellarators. Gyrokinetics is a good mathematical model of the drift-wave instabilities that cause turbulence. Multiple gyrokinetic turbulence codes that had great success comparing to tokamak experiments are being converted for use with stellarator geometry. This thesis describes such adaptations of the gyrokinetic turbulence code, GS2. Herein a new computational grid generator and upgrades to GS2 itself are described, tested, and benchmarked against three other gyrokinetic codes. Using GS2, detailed linear studies using the National Compact Stellarator Experiment (NCSX) geometry were conducted. The first compares stability in two equilibria with different β=(plasma pressure)/(magnetic pressure). Overall, the higher β case was more stable than the lower β case. As high β is important for MFE experiments, this is encouraging. The second compares NCSX linear stability to a tokamak case. NCSX was more stable with a 20% higher critical temperature gradient normalized by the minor radius, suggesting that the fusion power might be enhanced by ˜ 50%. In addition, the first nonlinear, non-axisymmetric GS2 simulations are presented. Finally, linear stability of two locations in a W7-AS plasma were compared. The experimentally-measured parameters used were from a W7-AS shot in which measured heat fluxes match neoclassical theory predictions at inner radii, but are too large for neoclassical predictions at outer radii. Results from GS2 linear simulations show that the outer location has higher gyrokinetic instability growth rates than at the inner one. Mixing-length estimates of the heat flux are within a factor of 3 of the experimental measurements, indicating that gyrokinetic turbulence may be responsible for the higher transport measured by the experiment in the outer regions. Future nonlinear simulations can explore this question in more detail. This work is supported by the Princeton Plasma Physics Laboratory, which is operated by Princeton University for the U.S. Department of Energy under Contract No. DE-AC02-09CH11466, and the SciDAC Center for the Study of Plasma Microturbulence.

A short introduction to the status and motivation for reversed field pinch and compact toroid research

NASA Astrophysics Data System (ADS)

Dreicer, H.

1987-09-01

Potential commercial fusion power systems must be acceptable from a safety and environmental standpoint. They must also promise to be competitive with other sources of energy (i.e., fossil, fission, etc.) when considered from the standpoint of the cost of electricity (COE) and the unit direct cost (UDC) in dollars/kWe. These costs are affected by a host of factors including recirculating power, plant availability, construction time, capital cost, etc., and are influenced by technological complexity. In an attempt to meet these requirements, the emphasis of fusion research in the United States has been moving toward smaller, lower-cost systems. There is increased interest in higher beta tokamaks and stellarators, and in compact alternate concepts such as the Reversed Field Pinch (RFP) and the Compact Toroids (CTs) which are, in part, the subject of this paper.

Compact ECEI system with in-vessel reflective optics for WEST.

PubMed

Nam, Y B; Park, H K; Lee, W; Yun, G S; Kim, M; Sabot, R; Elbeze, D; Lotte, P; Shen, J

2016-11-01

An electron cyclotron emission imaging (ECEI) diagnostic system for WEST (W Environment for Steady state Tokamak) is under development to study the MHD instabilities affected by tungsten impurities. The system will provide 2-D T e fluctuation images (width × height = ∼18 cm × ∼ 34 cm at low field side and ∼13 cm × ∼ 39 cm at high field side) from a poloidal cross section with high spatial (≤1.7 cm) and temporal (≤2 μs) resolutions. While the key concept and electronic structure are similar to that of prior ECEI systems on other tokamak devices such as KSTAR, DIII-D, or ASDEX-U, part of the imaging optics have to be placed inside the vacuum vessel in order to resolve issues on limited installation space and longer beam path to the detector position. The in-vessel optics consisting of two large curvature-radius mirrors are expected to withstand the extreme heating on long-pulse operation scenario (∼1000 s). The out-vessel optical housing is constructed as compact as possible to remove easily from the installation site in case of necessity. Commissioning of the system is scheduled on the second experimental WEST campaign end of 2017.

Overview, Progress, and Plans for the Compact Toroidal Hybrid Experiment

NASA Astrophysics Data System (ADS)

Hartwell, G. J.; Allen, N. R.; Ennis, D. A.; Hanson, J. D.; Howell, E. C.; Johnson, C. A.; Knowlton, S. F.; Kring, J. D.; Ma, X.; Maurer, D. A.; Ross, K. G.; Schmitt, J. C.; Traverso, P. J.; Williamson, E. N.

2017-10-01

The Compact Toroidal Hybrid (CTH) is an l = 2 , m = 5 torsatron/tokamak hybrid (R0 = 0.75 m, ap 0.2 m, and | B | <= 0.7 T) which generates highly configurable confining magnetic fields solely with external coils but typically uses up to 80 kA of plasma current for heating and disruption studies. The main goals of the CTH experiment are to study disruptive behavior as a function of applied 3D magnetic shaping, and to test and advance the V3FIT reconstruction code and NIMROD modeling of CTH. The disruptive density limit is observed to exceed the Greenwald limit as the vacuum transform is increased with no observed threshold for avoidance. Low-q operations (1.1 < q(a) < 2.0) are routine, with disruptions ceasing if the vacuum transform is raised above 0.07. Sawteeth are observed in CTH and have a similar phenomenology to tokamak sawteeth despite employing a 3D confining field. Application of vacuum transform has been demonstrated to reduce and eliminate the vertical drift of elongated discharges. Internal SXR diagnostics, in conjunction with external magnetics, extend the range of reconstruction accuracy into the plasma core. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

Full orbit computations of ripple-induced fusion {alpha}-particle losses from burning tokamak plasmas

DOE Office of Scientific and Technical Information (OSTI.GOV)

McClements, K.G.

A full orbit code is used to compute collisionless losses of fusion {alpha} particles from three proposed burning plasma tokamaks: the International Tokamak Experimental Reactor (ITER); a spherical tokamak power plant (STPP) [T. C. Hender, A. Bond, J. Edwards, P. J. Karditsas, K. G. McClements, J. Mustoe, D. V. Sherwood, G. M. Voss, and H. R. Wilson, Fusion Eng. Des. 48, 255 (2000)]; and a spherical tokamak components test facility (CTF) [H. R. Wilson, G. M. Voss, R. J. Akers, L. Appel, A. Dnestrovskij, O. Keating, T. C. Hender, M. J. Hole, G. Huysmans, A. Kirk, P. J. Knight, M.more » Loughlin, K. G. McClements, M. R. O'Brien, and D. Yu. Sychugov, Proceedings of the 20th IAEA Fusion Energy Conference, Invited Paper FT/3-1Ra]. It has been suggested that {alpha} particle transport could be enhanced due to cyclotron resonance with the toroidal magnetic field ripple. However, calculations for inductive operation in ITER yield a loss rate that appears to be broadly consistent with the predictions of guiding center theory, falling monotonically as the number of toroidal field coils N is increased (and hence the ripple amplitude is decreased). For STPP and CTF the loss rate does not decrease monotonically with N, but collisionless losses are generally low in absolute terms. As in the case of ITER, there is no evidence that finite Larmor radius effects would seriously degrade fusion {alpha}-particle confinement.« less

PIE on Safety-Tested Loose Particles from Irradiated Compact 4-4-2

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Gerczak, Tyler J.; Morris, Robert Noel

2016-04-01

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High Temperature Gas-cooled Reactors (HTGRs). This work is sponsored by the Department of Energy Office of Nuclear Energy (DOE-NE) through the Advanced Reactor Technologies (ART) Office under the Advanced Gas Reactor Fuel Development and Qualification (AGR) Program. The AGR-1 experiment was the first in a series of TRISO fuel irradiation tests initiated in 2006. The AGR-1 TRISO particles and fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 using laboratory-scale equipment and irradiated for 3 years in themore » Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. Post-irradiation examination was performed at INL and ORNL to study how the fuel behaved during irradiation, and to test fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing and post-safety testing PIE conducted at ORNL on loose particles extracted from irradiated AGR-1 Compact 4-4-2.« less

Largescale Long-term particle Simulations of Runaway electrons in Tokamaks

NASA Astrophysics Data System (ADS)

Liu, Jian; Qin, Hong; Wang, Yulei

2016-10-01

To understand runaway dynamical behavior is crucial to assess the safety of tokamaks. Though many important analytical and numerical results have been achieved, the overall dynamic behaviors of runaway electrons in a realistic tokamak configuration is still rather vague. In this work, the secular full-orbit simulations of runaway electrons are carried out based on a relativistic volume-preserving algorithm. Detailed phase-space behaviors of runaway electrons are investigated in different timescales spanning 11 orders. A detailed analysis of the collisionless neoclassical scattering is provided when considering the coupling between the rotation of momentum vector and the background field. In large timescale, the initial condition of runaway electrons in phase space globally influences the runaway distribution. It is discovered that parameters and field configuration of tokamaks can modify the runaway electron dynamics significantly. Simulations on 10 million cores of supercomputer using the APT code have been completed. A resolution of 107 in phase space is used, and simulations are performed for 1011 time steps. Largescale simulations show that in a realistic fusion reactor, the concern of runaway electrons is not as serious as previously thought. This research was supported by National Magnetic Connement Fusion Energy Research Project (2015GB111003, 2014GB124005), the National Natural Science Foundation of China (NSFC-11575185, 11575186) and the GeoAlgorithmic Plasma Simulator (GAPS) Project.

Development of plasma bolometers using fiber-optic temperature sensors

NASA Astrophysics Data System (ADS)

Reinke, M. L.; Han, M.; Liu, G.; van Eden, G. G.; Evenblij, R.; Haverdings, M.; Stratton, B. C.

2016-11-01

Measurements of radiated power in magnetically confined plasmas are important for exhaust studies in present experiments and expected to be a critical diagnostic for future fusion reactors. Resistive bolometer sensors have long been utilized in tokamaks and helical devices but suffer from electromagnetic interference (EMI). Results are shown from initial testing of a new bolometer concept based on fiber-optic temperature sensor technology. A small, 80 μm diameter, 200 μm long silicon pillar attached to the end of a single mode fiber-optic cable acts as a Fabry-Pérot cavity when broadband light, λo ˜ 1550 nm, is transmitted along the fiber. Changes in temperature alter the optical path length of the cavity primarily through the thermo-optic effect, resulting in a shift of fringes reflected from the pillar detected using an I-MON 512 OEM spectrometer. While initially designed for use in liquids, this sensor has ideal properties for use as a plasma bolometer: a time constant, in air, of ˜150 ms, strong absorption in the spectral range of plasma emission, immunity to local EMI, and the ability to measure changes in temperature remotely. Its compact design offers unique opportunities for integration into the vacuum environment in places unsuitable for a resistive bolometer. Using a variable focus 5 mW, 405 nm, modulating laser, the signal to noise ratio versus power density of various bolometer technologies are directly compared, estimating the noise equivalent power density (NEPD). Present tests show the fiber-optic bolometer to have NEPD of 5-10 W/m2 when compared to those of the resistive bolometer which can achieve <0.5 W/m2 in the laboratory, but this can degrade to 1-2 W/m2 or worse when installed on a tokamak. Concepts are discussed to improve the signal to noise ratio of this new fiber-optic bolometer by reducing the pillar height and adding thin metallic coatings, along with improving the spectral resolution of the interrogator.

Studies of PuF sub 6 and transplutonic materials' critical properties for space high power nuclear pumped lasers

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gu, A.G.; Miller, M.S.

1991-01-01

All space missions require a reliable, compact source of energy. This paper describes preliminary neutronics studies of pocket'' reactor concepts employing PuF{sub 6} and transplutonic materials as fuels for space high power/energy Nuclear Pumped Lasers (NPLs). Previous research has studied NPL reactor concepts with thin fuel layers, aerosol fuels and gaseous UF{sub 6}. The total reactor volumes for compact reactors with these types of fuels typically range from 3 m{sup 3} to 50 m{sup 3}. By employing PuF{sub 6} and transplutonic fuels at the same low densities, a calculated value for Keff of 1.2 has been achieved for conditions ofmore » 900 K and 5 atm, with total reactor volumes of 1.5 m{sup 3} for PuF{sub 6}, 0.51 m{sup 3} for Am-242m, 0.58 m{sup 3} for Cm-245 and 0.63 m{sup 3} for Cf-249.« less

Impact of physics and technology innovations on compact tokamak fusion pilot plants

NASA Astrophysics Data System (ADS)

Menard, Jonathan

2016-10-01

For magnetic fusion to be economically attractive and have near-term impact on the world energy scene it is important to focus on key physics and technology innovations that could enable net electricity production at reduced size and cost. The tokamak is presently closest to achieving the fusion conditions necessary for net electricity at acceptable device size, although sustaining high-performance scenarios free of disruptions remains a significant challenge for the tokamak approach. Previous pilot plant studies have shown that electricity gain is proportional to the product of the fusion gain, blanket thermal conversion efficiency, and auxiliary heating wall-plug efficiency. In this work, the impact of several innovations is assessed with respect to maximizing fusion gain. At fixed bootstrap current fraction, fusion gain varies approximately as the square of the confinement multiplier, normalized beta, and major radius, and varies as the toroidal field and elongation both to the third power. For example, REBCO high-temperature superconductors (HTS) offer the potential to operate at much higher toroidal field than present fusion magnets, but HTS cables are also beginning to access winding pack current densities up to an order of magnitude higher than present technology, and smaller HTS TF magnet sizes make low-aspect-ratio HTS tokamaks potentially attractive by leveraging naturally higher normalized beta and elongation. Further, advances in kinetic stabilization and feedback control of resistive wall modes could also enable significant increases in normalized beta and fusion gain. Significant reductions in pilot plant size will also likely require increased plasma energy confinement, and control of turbulence and/or low edge recycling (for example using lithium walls) would have major impact on fusion gain. Reduced device size could also exacerbate divertor heat loads, and the impact of novel divertor solutions on pilot plant configurations is addressed. For missions including tritium breeding, high-thermal-efficiency liquid metal breeding blankets are attractive, and novel immersion blankets offer the potential for simplified fabrication and maintenance and reduced cost. Lastly, the optimal aspect ratio for a tokamak pilot plant is likely a function of the device mission and associated cost, with low aspect ratio favored for minimizing TF magnet mass and higher aspect ratio favored for minimizing blanket mass. The interplay between a range of physics and technology innovations for enabling compact pilot plants will be described. This work was supported by U.S. DOE Contract Number DE-AC02-09CH11466.

Design of megawatt power level heat pipe reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mcclure, Patrick Ray; Poston, David Irvin; Dasari, Venkateswara Rao

An important niche for nuclear energy is the need for power at remote locations removed from a reliable electrical grid. Nuclear energy has potential applications at strategic defense locations, theaters of battle, remote communities, and emergency locations. With proper safeguards, a 1 to 10-MWe (megawatt electric) mobile reactor system could provide robust, self-contained, and long-term power in any environment. Heat pipe-cooled fast-spectrum nuclear reactors have been identified as a candidate for these applications. Heat pipe reactors, using alkali metal heat pipes, are perfectly suited for mobile applications because their nature is inherently simpler, smaller, and more reliable than “traditional” reactors.more » The goal of this project was to develop a scalable conceptual design for a compact reactor and to identify scaling issues for compact heat pipe cooled reactors in general. Toward this goal two detailed concepts were developed, the first concept with more conventional materials and a power of about 2 MWe and a the second concept with less conventional materials and a power level of about 5 MWe. A series of more qualitative advanced designs were developed (with less detail) that show power levels can be pushed to approximately 30 MWe.« less

Single Null Negative Triangularity Tokamak for Power Handling

NASA Astrophysics Data System (ADS)

Kikuchi, Mitsuru; Medvedev, S.; Takizuka, T.; Sauter, O.; Merle, A.; Coda, S.; Chen, D.; Li, J. X.

2017-10-01

Power and particle control in fusion reactor is challenge and we proposed the negative triangularity tokamak (NTT) to eliminate ELM by operating L-mode edge with improved core confinement. The SN configuration has more flexibility in shaping by adopting rectangular-shaped TF coils. The limiting normalized beta is 3.56 with wall stabilization and 3.14 without wall. The vertical stability is assured under a reasonable control system. The wetted area on the divertor plates becomes wider in proportion to the larger major radius at the divertor strike points due to the NT configuration. In addition to the major-radius effect, the ``Flux Tune Expansion (FTE)'' is adopted to further reduce the heat load on the divertor plate by factor of 2.6 with a coil current 3 MA. L-mode edge also allows further increase in wetted area. The fusion power of 3 GW is deliverable only at normalized beta 2.1. Therefore this reactor may be operable stably against the serious MHD activities. The CD power for SS operation is 175 MW at Q = 17. AC operation is also possible option. A required HH factor is relatively modest H = 1.12.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gorelenkov, Nikolai N

The area of energetic particle (EP) physics of fusion research has been actively and extensively researched in recent decades. The progress achieved in advancing and understanding EP physics has been substantial since the last comprehensive review on this topic by W.W. Heidbrink and G.J. Sadler [1]. That review coincided with the start of deuterium-tritium (DT) experiments on Tokamak Fusion Test reactor (TFTR) and full scale fusion alphas physics studies. Fusion research in recent years has been influenced by EP physics in many ways including the limitations imposed by the "sea" of Alfven eigenmodes (AE) in particular by the toroidicityinduced AEsmore » (TAE) modes and reversed shear Alfven (RSAE). In present paper we attempt a broad review of EP physics progress in tokamaks and spherical tori since the first DT experiments on TFTR and JET (Joint European Torus) including helical/stellarator devices. Introductory discussions on basic ingredients of EP physics, i.e. particle orbits in STs, fundamental diagnostic techniques of EPs and instabilities, wave particle resonances and others are given to help understanding the advanced topics of EP physics. At the end we cover important and interesting physics issues toward the burning plasma experiments such as ITER (International Thermonuclear Experimental Reactor).« less

High Power LaB6 Plasma Source Performance for the Lockheed Martin Compact Fusion Reactor Experiment

NASA Astrophysics Data System (ADS)

Heinrich, Jonathon

2016-10-01

Lockheed Martin's Compact Fusion Reactor (CFR) concept is a linear encapsulated ring cusp. Due to the complex field geometry, plasma injection into the device requires careful consideration. A high power thermionic plasma source (>0.25MW; >10A/cm2) has been developed with consideration to phase space for optimal coupling. We present the performance of the plasma source, comparison with alternative plasma sources, and plasma coupling with the CFR field configuration. ©2016 Lockheed Martin Corporation. All Rights Reserved.

Molecular Dynamics Simulation of Hydrogen Trapping on Sigma 5 Tungsten Grain Boundaries

NASA Astrophysics Data System (ADS)

Al-Shalash, Aws Mohammed Taha

Tungsten as a plasma facing material is the predominant contender for future Tokamak reactor environments. The interaction between the plasma particles and tungsten is crucial to be studied for successful usage and design of tungsten in the plasma facing components ensuring the reliability and longevity of the fusion reactors. The bombardment of the sigma 5 polycrystalline tungsten was modeled using the molecular dynamics simulation through the large-scale atomic/molecular massively parallel simulator (LAMMPS) code and Tersoff type interatomic potential. By simulating the operational conditions of the Tokamak reactors, the hydrogen trapping rate, implantation distribution, and bubble formation was investigated at various temperatures (300-1200 K) and various hydrogen incident energy (20-100 eV). The substrate's temperature increases the deflected H atoms, and increases the penetration depth for the ones that go through. As well, the lower temperature tungsten substrates retain more H atoms. Increasing the bombarded hydrogen's energy increases the trapping and retention rate and the depth of penetration. Another experiments were conducted to determine whether the Sigma5 grain boundary's (GB) location affects the trapping profiles in H. The findings are ranges from small effect on deflection rates at low H energies to no effect at high H energies. However, there is a considerable effect on shifting the trapping depth profile upward toward the surface when raising the GB closer to the surface. Hydrogen atoms are highly mobile on tungsten substrate, yet no bubble formation was witnessed.

BESAFE II: Accident safety analysis code for MFE reactor designs

NASA Astrophysics Data System (ADS)

Sevigny, Lawrence Michael

The viability of controlled thermonuclear fusion as an alternative energy source hinges on its desirability from an economic and an environmental and safety standpoint. It is the latter which is the focus of this thesis. For magnetic fusion energy (MFE) devices, the safety concerns equate to a design's behavior during a worst-case accident scenario which is the loss of coolant accident (LOCA). In this dissertation, we examine the behavior of MFE devices during a LOCA and how this behavior relates to the safety characteristics of the machine; in particular the acute, whole-body, early dose. In doing so, we have produced an accident safety code, BESAFE II, now available to the fusion reactor design community. The Appendix constitutes the User's Manual for BESAFE II. The theory behind early dose calculations including the mobilization of activation products is presented in Chapter 2. Since mobilization of activation products is a strong function of temperature, it becomes necessary to calculate the thermal response of a design during a LOCA in order to determine the fraction of the activation products which are mobilized and thus become the source for the dose. The code BESAFE II is designed to determine the temperature history of each region of a design and determine the resulting mobilization of activation products at each point in time during the LOCA. The BESAFE II methodology is discussed in Chapter 4, followed by demonstrations of its use for two reference design cases: a PCA-Li tokamak and a SiC-He tokamak. Of these two cases, it is shown that the SiC-He tokamak is a better design from an accident safety standpoint than the PCA-Li tokamak. It is also found that doses derived from temperature-dependent mobilization data are different than those predicted using set mobilization categories such as those that involve Piet fractions. This demonstrates the need for more experimental data on fusion materials. The possibility for future improvements and modifications to BESAFE II is discussed in Chapter 6, for example, by adding additional environmental indices such as a waste disposal index. The biggest improvement to BESAFE II would be an increase in the database of activation product mobilization for a larger spectrum of fusion reactor materials. The ultimate goal we have is for BESAFE II to become part of a systems design program which would include economic factors and allow both safety and the cost of electricity to influence design.

Li experiments at the tokamak T-11 M in field of steady state PFC investigations

NASA Astrophysics Data System (ADS)

Mirnov, S. V.; Lazarev, V. B.

2011-08-01

The renewable plasma facing components (PFCs) of steady state tokamak-reactor can be created in framework of Lithium emitter-collector concept, which suggests Li-loop development close the Li-PFC and plasma periphery. It should ensure: Li-emission from PFC into the plasma, plasma periphery cooling by non-coronal Li radiation, Li ions collection before their loss on the wall and Li return into emitter. The subjects of the last T-11 M investigations were the Lithium collection by limiters and Lithium removal from the wall during tokamak conditioning. The Lithium behavior was studied with witness samples and mobile graphite probe. It was shown that Li-deposit on the sides of rail Li-limiter (collector) is proportional to the Li-emission from the Li-limiter (emitter). Lithium deposit on the ion-drift side of Li-limiter is up to 2-3 times more than on the electron-side. The efficiency of Li-collection by T-11 M limiters can be 60 ± 20% of total Lithium emission from Li-limiter during plasma discharges.

Energetic particle-driven compressional Alfvén eigenmodes and prospects for ion cyclotron emission studies in fusion plasmas

DOE PAGES

Gorelenkov, N. N.

2016-10-01

As a fundamental plasma oscillation the compressional Alfvén waves (CAW) are interesting for plasma scientists both academically and in applications for fusion plasmas. They are believed to be responsible for the ion cyclotron emission (ICE) observed in many tokamaks. The theory of CAW and ICE was significantly advanced at the end of 20th century in particular motivated by first DT experiments on TFTR and subsequent JET DT experimental studies. More recently, ICE theory was advanced by ST (or spherical torus) experiments with the detailed theoretical and experimental studies of the properties of each instability signal. There the instability responsible formore » ICE signals previously indistinguishable in high aspect ratio tokamaks became the subjects of experimental studies. We discuss further the prospects of ICE theory and its applications for future burning plasma (BP) experiments such as the ITER tokamak-reactor prototype being build in France where neutrons and gamma rays escaping the plasma create extremely challenging conditions for fusion alpha particle diagnostics.a« less

Tokamak Operation with Safety Factor q 95 < 2 via Control of MHD Stability

DOE PAGES

Piovesan, Paolo; Hanson, Jeremy M.; Martin, Piero; ...

2014-07-24

Magnetic feedback control of the resistive-wall mode has enabled DIII-D to access stable operation at safety factor q95 = 1:9 in divertor plasmas for 150 instability growth times. Magnetohydrodynamic stability sets a hard, disruptive limit on the minimum edge safety factor achievable in a tokamak, or on the maximum plasma current at given toroidal magnetic eld. In tokamaks with a divertor, the limit occurs at q95 = 2, as con rmed in DIII-D. Since the energy con cement time scales linearly with current, this also bounds the performance of a fusion reactor. DIII-D has overcome this limit, opening a wholemore » new high-current regime not accessible before. This result brings signi cant possible bene ts in terms of fusion performance, but it also extends resistive wall mode physics and its control to conditions never explored before. In present experiments, q95 < 2 operation is eventually halted by voltage limits reached in the feedback power supplies, not by intrinsic physics issues. Improvements to power supplies and to control algorithms have the potential to further extend this regime.« less

Interactive Plasma Physics Education Using Data from Fusion Experiments

NASA Astrophysics Data System (ADS)

Calderon, Brisa; Davis, Bill; Zwicker, Andrew

2010-11-01

The Internet Plasma Physics Education Experience (IPPEX) website was created in 1996 to give users access to data from plasma and fusion experiments. Interactive material on electricity, magnetism, matter, and energy was presented to generate interest and prepare users to understand data from a fusion experiment. Initially, users were allowed to analyze real-time and archival data from the Tokamak Fusion Test Reactor (TFTR) experiment. IPPEX won numerous awards for its novel approach of allowing users to participate in ongoing research. However, the latest revisions of IPPEX were in 2001 and the interactive material is no longer functional on modern browsers. Also, access to real-time data was lost when TFTR was shut down. The interactive material on IPPEX is being rewritten in ActionScript3.0, and real-time and archival data from the National Spherical Tokamak Experiment (NSTX) will be made available to users. New tools like EFIT animations, fast cameras, and plots of important plasma parameters will be included along with an existing Java-based ``virtual tokamak.'' Screenshots from the upgraded website and future directions will be presented.

The Science and Technology Case for High-Field Fusion

NASA Astrophysics Data System (ADS)

Whyte, D.

2017-10-01

This review will focus on the origin, development and new opportunities of a strategy for fusion energy based on the high-field approach. In this approach confinement devices are designed at the maximum possible value of vacuum magnetic field strength, B. The integrated electrical, mechanical and cooling engineering challenges of high-field on coil (Bcoil) , large-bore electromagnets are examined for both copper and superconductor materials. These engineering challenges are confronted because of the profound science advantages provided by high-B, which are derived and reviewed: high fusion power density, B4, in compact devices, thermonuclear plasmas with significant stability margin, and, in tokamaks, access to higher plasma density. Two distinct high-field strategies emerged in the 1980's. The first was compact, cryogenically-cooled copper devices (BPX, IGNITOR, FIRE) with Bcoil>20 T, while the second was a large-volume, Nb3Sn superconductor device with Bcoil <12 T; with the second path exclusively chosen ca. 2000 with the ITER construction decision. The reasoning, advantages and challenges of that decision are discussed. Yet since that decision, a new opportunity has arisen: compact, Rare Earth Barium Copper Oxide (REBCO) superconductor-based devices with Bcoil >20 T; a strategy that essentially combines the best components of the two previous strategies. Recent activities examining the technology and science implications of this new strategy are reviewed. On the technology side, REBCO superconductors have now been used to produce Bcoil>40 T in small-bore electromagnets, enabled by rapid progress in manufactured REBCO conductor quality, coil modularity and flexible operating temperature range. Specific tokamak designs, over a range of aspect ratios, have been developed to take scientific advantage of these features in various ways, and will be described.

Rigorous-two-Steps scheme of TRIPOLI-4® Monte Carlo code validation for shutdown dose rate calculation

NASA Astrophysics Data System (ADS)

Jaboulay, Jean-Charles; Brun, Emeric; Hugot, François-Xavier; Huynh, Tan-Dat; Malouch, Fadhel; Mancusi, Davide; Tsilanizara, Aime

2017-09-01

After fission or fusion reactor shutdown the activated structure emits decay photons. For maintenance operations the radiation dose map must be established in the reactor building. Several calculation schemes have been developed to calculate the shutdown dose rate. These schemes are widely developed in fusion application and more precisely for the ITER tokamak. This paper presents the rigorous-two-steps scheme implemented at CEA. It is based on the TRIPOLI-4® Monte Carlo code and the inventory code MENDEL. The ITER shutdown dose rate benchmark has been carried out, results are in a good agreement with the other participant.

ITER activities and fusion technology

NASA Astrophysics Data System (ADS)

Seki, M.

2007-10-01

At the 21st IAEA Fusion Energy Conference, 68 and 67 papers were presented in the categories of ITER activities and fusion technology, respectively. ITER performance prediction, results of technology R&D and the construction preparation provide good confidence in ITER realization. The superconducting tokamak EAST achieved the first plasma just before the conference. The construction of other new experimental machines has also shown steady progress. Future reactor studies stress the importance of down sizing and a steady-state approach. Reactor technology in the field of blanket including the ITER TBM programme and materials for the demonstration power plant showed sound progress in both R&D and design activities.

Retarding field energy analyzer for the Saskatchewan Torus-Modified plasma boundary

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dreval, M.; Rohraff, D.; Xiao, C.

2009-10-15

The retarding field energy analyzer (RFA) is a simple and reliable diagnostic technique to measure the ion temperature in the scrape-off layer and edge of magnetic fusion devices. Design and operation features of a single-sided (facing the ion flow) RFA for ion temperature measurements in the Saskatchewan Torus-Modified (STOR-M) tokamak are described. Its compact size (21x15x20 mm{sup 3}) allows RFA measurements without perturbing plasma significantly. Both ion and electron temperature have been measured by RFA in the STOR-M tokamak. A method is proposed to correct the effects of ion flow on the ion temperature using the simultaneously measured Mach number.more » The measured electron temperature is consistent with the previously reported Langmuir probe data. Abnormal behavior of the RFA has been observed in both ion and electron modes when RFA is inserted deep into the plasma.« less

METHOD AND APPARATUS FOR REACTOR SAFETY CONTROL

DOEpatents

Huston, N.E.

1961-06-01

A self-contained nuclear reactor fuse controlled device tron absorbing material, normally in a compact form but which can be expanded into an extended form presenting a large surface for neutron absorption when triggered by an increase in neutron flux, is described.

PIE on Safety-Tested AGR-1 Compact 5-1-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Morris, Robert Noel; Baldwin, Charles A.

Post-irradiation examination (PIE) is being performed in support of tristructural isotropic (TRISO) coated particle fuel development and qualification for High-Temperature Gas-cooled Reactors (HTGRs). AGR-1 was the first in a series of TRISO fuel irradiation experiments initiated in 2006 under the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program; this work continues to be funded by the Department of Energy's Office of Nuclear Energy as part of the Advanced Reactor Technologies (ART) initiative. AGR-1 fuel compacts were fabricated at Oak Ridge National Laboratory (ORNL) in 2006 and irradiated for three years in the Idaho National Laboratory (INL) Advanced Test Reactormore » (ATR) to demonstrate and evaluate fuel performance under HTGR irradiation conditions. PIE is being performed at INL and ORNL to study how the fuel behaved during irradiation, and to examine fuel performance during exposure to elevated temperatures at or above temperatures that could occur during a depressurized conduction cooldown event. This report summarizes safety testing of irradiated AGR-1 Compact 5-1-1 in the ORNL Core Conduction Cooldown Test Facility (CCCTF) and post-safety testing PIE.« less

Book Review:

NASA Astrophysics Data System (ADS)

Rogister, A. L.

2004-03-01

John Wesson’s well known book, now re-edited for the third time, provides an excellent introduction to fusion oriented plasma physics in tokamaks. The author’s task was a very challenging one, for a confined plasma is a complex system characterised by a variety of dimensionless parameters and its properties change qualitatively when certain threshold values are reached in this multi-parameter space. As a consequence, theoretical description is required at different levels, which are complementary: particle orbits, kinetic and fluid descriptions, but also intuitive and empirical approaches. Theory must be carried out on many fronts: equilibrium, instabilities, heating, transport etc. Since the properties of the confined plasma depend on the boundary conditions, the physics of plasmas along open magnetic field lines and plasma surface interaction processes must also be accounted for. Those subjects (and others) are discussed in depth in chapters 2 9. Chapter 1 mostly deals with ignition requirements and the tokamak concept, while chapter 14 provides a list of useful relations: differential operators, collision times, characteristic lengths and frequencies, expressions for the neoclassical resistivity and heat conduction, the bootstrap current etc. The presentation is sufficiently broad and thorough that specialists within tokamak research can either pick useful and up-to-date information or find an authoritative introduction into other areas of the subject. It is also clear and concise so that it should provide an attractive and accurate initiation for those wishing to enter the field and for outsiders who would like to understand the concepts and be informed about the goals and challenges on the horizon. Validation of theoretical models requires adequately resolved experimental data for the various equilibrium profiles (clearly a challenge in the vicinity of transport barriers) and the fluctuations to which instabilities give rise. Chapter 10 is therefore devoted to an introduction to diagnostics for tokamaks. The complexity of fusion plasmas is attested to by the discovery of new phenomena and new operational regimes as machine size and power increased and the diagnostic tools improved over the forty years of research on magnetic confinement. The history of those discoveries in the devices which have been built worldwide after the results obtained on the first tokamaks at the Kurchatov Institute had been confirmed is outlined in chapters 11 12. Particular emphasis is naturally given to the results from the larger tokamaks: ASDEX Upgrade, DIII-D, TFTR, JT-60/JT-60U and JET. Chapter 13 is devoted to the International Tokamak Experimental Reactor and prospects beyond ITER. Examples of operational regimes and of often unexpected phenomena are the linear and saturated ohmic confinement modes, confinement degradation when auxiliary heating is applied, the high energy confinement mode, the formation of internal transport barriers in weak or negative central shear discharges, sawtooth relaxations, disruptions, multifaceted asymmetric radiation from the edge, edge localised modes, etc. The relevant observations are described very thoroughly with the support of numerous selected figures and their physical interpretation, a major topic of the book, is carefully discussed on the basis of simplified but convincing mathematical models. With respect to the previous edition (1997), a few additions have been introduced; those concern plasma rotation (section 3.13), internal transport barriers (4.14), the role of radial electric field shear (4.19), turbulence simulations (4.21), impurity transport (4.22) and neoclassical drive of tearing modes (7.3). It is my personal feeling that some of those additions should have been somewhat more elaborated. A few pages have finally been added concerning the TCV, START, MAST, NSTX and ASDEX Upgrade tokamaks. With this book, John Wesson offers the fusion community a very precious and thorough survey of tokamak physics, from basic principles to interpretation of experimental data, and to a wider readership an elegant and authoritative introduction to the challenges that are associated with the development of the tokamak reactor, a source of limitless and clean thermonuclear power. This reference book should be on the shelf of every fusion scientist and graduate student.

Heat pulse propagation studies on DIII-D and the Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Fredrickson, E. D.; Austin, M. E.; Groebner, R.; Manickam, J.; Rice, B.; Schmidt, G.; Snider, R.

2000-12-01

Sawtooth phenomena have been studied on DIII-D and the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24]. In the experiments the sawtooth characteristics were studied with fast electron temperature (ECE) and soft x-ray diagnostics. For the first time, measurements of a strong ballistic electron heat pulse were made in a shaped tokamak (DIII-D) [J. Luxon and DIII-D Group, in Proceedings of the 11th International Conference on Plasma Physics and Controlled Nuclear Fusion Research, Kyoto (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 159] and the "ballistic effect" was stronger than was previously reported on TFTR. Evidence is presented in this paper that the ballistic effect is related to the fast growth phase of the sawtooth precursor. Fast, 2 ms interval, measurements on DIII-D were made of the ion temperature evolution following sawteeth and partial sawteeth to document the ion heat pulse characteristics. It is found that the ion heat pulse does not exhibit the very fast, "ballistic" behavior seen for the electrons. Further, for the first time it is shown that the electron heat pulses from partial sawtooth crashes (on DIII-D and TFTR) are seen to propagate at speeds close to those expected from the power balance calculations of the thermal diffusivities whereas heat pulses from fishbones propagate at rates more consistent with sawtooth induced heat pulses. These results suggest that the fast propagation of sawtooth-induced heat pulses is not a feature of nonlinear transport models, but that magnetohydrodynamic events can have a strong effect on electron thermal transport.

Advanced Fuels Reactor using Aneutronic Rodless Ultra Low Aspect Ratio Tokamak Hydrogenic Plasmas

NASA Astrophysics Data System (ADS)

Ribeiro, Celso

2015-11-01

The use of advanced fuels for fusion reactor is conventionally envisaged for field reversed configuration (FRC) devices. It is proposed here a preliminary study about the use of these fuels but on an aneutronic Rodless Ultra Low Aspect Ratio (RULART) hydrogenic plasmas. The idea is to inject micro-size boron pellets vertically at the inboard side (HFS, where TF is very high and the tokamak electron temperature is relatively low because of profile), synchronised with a proton NBI pointed to this region. Therefore, p-B reactions should occur and alpha particles produced. These pellets will act as an edge-like disturbance only (cp. killer pellet, although the vertical HFS should make this less critical, since the unablated part should appear in the bottom of the device). The boron cloud will appear at midplance, possibly as a MARFE-look like. Scaling of the p-B reactions by varying the NBI energy should be compared with the predictions of nuclear physics. This could be an alternative to the FRC approach, without the difficulties of the optimization of the FRC low confinement time. Instead, a robust good tokamak confinement with high local HFS TF (enhanced due to the ultra low aspect ratio and low pitch angle) is used. The plasma central post makes the RULART concept attractive because of the proximity of NBI path and also because a fraction of born alphas will cross the plasma post and dragged into it in the direction of the central plasma post current, escaping vertically into a hole in the bias plate and reaching the direct electricity converter, such as in the FRC concept.

The external kink mode in diverted tokamaks

NASA Astrophysics Data System (ADS)

Turnbull, A. D.; Hanson, J. M.; Turco, F.; Ferraro, N. M.; Lanctot, M. J.; Lao, L. L.; Strait, E. J.; Piovesan, P.; Martin, P.

2016-06-01

> . The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.

Design and Implementation of a 200kW, 28GHz gyrotron system for the Compact Toroidal Hybrid Experiment

NASA Astrophysics Data System (ADS)

Hartwell, G. J.; Knowlton, S. F.; Ennis, D. A.; Maurer, D. A.; Bigelow, T.

2016-10-01

The Compact Toroidal Hybrid (CTH) is an l = 2 , m = 5 torsatron/tokamak hybrid (R0 = 0.75 m, ap 0.2 m, and | B | <= 0.7 T). It can generate its highly configurable confining magnetic fields solely with external coils, but typically operates with up to 80 kA of ohmically-generated plasma current for heating. New studies of edge plasma transport in stellarator geometries will benefit from CTH operating as a pure torsatron with a high temperature edge plasma. Accordingly, a 28 GHz, 200 kW gyrotron operating at 2nd harmonic for ECRH is being installed to supplement the existing 15 kW klystron system operating at the fundamental frequency; the latter will be used to initially generate the plasma. Ray-tracing calculations that guide the selection of launching position, antenna focal length, and beam-steering characteristics of the ECRH have been performed with the TRAVIS code [ 1 ] . The calculated absorption is up to 95.7% for vertically propagating rays, however, the absorption is more sensitive to magnetic field variations than for a side launch where the field gradient is tokamak-like. The design of the waveguide path and components for the top-launch scenario will be presented. This work is supported by U.S. Department of Energy Grant No. DE-FG02-00ER54610.

Particle and momentum confinement in tokamak plasmas with unbalanced neutral beam injection and strong rotation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Malik, M.A.

1988-01-01

There is a self-consistent theory of the effects of neutral beam injection on impurity transport in tokamak plasmas. The theory predicts that co-injection drives impurities outward and that counter-injection enhances the normally inward flow of impurities. The theory was applied to carry out a detailed analysis of the large experimental database from the PLT and the ISX-B tokamaks. The theory was found to generally model the experimental data quite well. It is, therefore, concluded that neutral beam co-injection can drive impurities outward to achieve clean central plasmas and a cool radiating edge. Theoretical predictions for future thermonuclear reactors such asmore » INTOR, TIBER II, and ITER indicated that neutral beam driven flow reversal might be an effective impurity control method if the rate of beam momentum deposited per plasma ion is adequate. The external momentum drag, which is a pivotal concept in impurity flow reversal theory, is correctly predicted by the gyroviscous theory of momentum confinement. The theory was applied to analyze experimental data from the PLT and the PDX tokamaks with exact experimental conditions. The theory was found to be in excellent agreement with experiment over a wide range of parameters. It is, therefore, possible to formulate the impurity transport theory from first principles, without resort to empiricism.« less

Effects of Hot Limiter Biasing on Tokamak Runaway Discharges

NASA Astrophysics Data System (ADS)

Salar Elahi, A.; Ghoranneviss, M.; Ghanbari, M. R.

2013-10-01

In this research hot limiter biasing effects on the Runaway discharges were investigated. First wall of the tokamak reactors can affects serious damage due to the high energy runaway electrons during a major disruption and therefore its life time can be reduced. Therefore, it is important to find methods to decrease runaway electron generation and their energy. Tokamak limiter biasing is one of the methods for controlling the radial electric field and can induce a transition to an improved confinement state. In this article generation of runaway electrons and the energy they can obtain will be investigated theoretically. Moreover, in order to apply radial biasing an emissive limiter biasing is utilized. The biased limiter can apply +380 V in the status of cold and hot to the plasma and result in the increase of negative bias current in hot status. In fact, in this experiment we try to decrease the generation of runaway electrons and their energy by using emissive limiter biasing inserted on the IR-T1 tokamak. The mean energy of these electrons was obtained by spectroscopy of hard X-ray. Also, the plasma current center shift was measured from the vertical field coil characteristics in presence of limiter biasing. The calculation is made focusing on the vertical field coil current and voltage changes due to a horizontal displacement of plasma column.

Progress Toward Steady State Tokamak Operation Exploiting the high bootstrap current fraction regime

NASA Astrophysics Data System (ADS)

Ren, Q.

2015-11-01

Recent DIII-D experiments have advanced the normalized fusion performance of the high bootstrap current fraction tokamak regime toward reactor-relevant steady state operation. The experiments, conducted by a joint team of researchers from the DIII-D and EAST tokamaks, developed a fully noninductive scenario that could be extended on EAST to a demonstration of long pulse steady-state tokamak operation. Fully noninductive plasmas with extremely high values of the poloidal beta, βp >= 4 , have been sustained at βT >= 2 % for long durations with excellent energy confinement quality (H98y,2 >= 1 . 5) and internal transport barriers (ITBs) generated at large minor radius (>= 0 . 6) in all channels (Te, Ti, ne, VTf). Large bootstrap fraction (fBS ~ 80 %) has been obtained with high βp. ITBs have been shown to be compatible with steady state operation. Because of the unusually large ITB radius, normalized pressure is not limited to low βN values by internal ITB-driven modes. βN up to ~4.3 has been obtained by optimizing the plasma-wall distance. The scenario is robust against several variations, including replacing some on-axis with off-axis neutral beam injection (NBI), adding electron cyclotron (EC) heating, and reducing the NBI torque by a factor of 2. This latter observation is particularly promising for extension of the scenario to EAST, where maximum power is obtained with balanced NBI injection, and to a reactor, expected to have low rotation. However, modeling of this regime has provided new challenges to state-of-the-art modeling capabilities: quasilinear models can dramatically underpredict the electron transport, and the Sauter bootstrap current can be insufficient. The analysis shows first-principle NEO is in good agreement with experiments for the bootstrap current calculation and ETG modes with a larger saturated amplitude or EM modes may provide the missing electron transport. Work supported in part by the US DOE under DE-FC02-04ER54698, DE-AC52-07NA27344, DE-AC02-09CH11466, and the NMCFP of China under 2015GB110000 and 2015GB102000.

Magnetically-induced forces on a ferromagnetic HT-9 first wall/blanket module

NASA Astrophysics Data System (ADS)

Lechtenberg, T. A.; Dahms, C. F.; Attaya, H.

1984-05-01

A model of the Starfire commercial tokamak reactor was used as the basis for calculating magnetic loads induced on typical fusion reactor first wall components fabricated of ferromagnetic material. The component analyzed was the first wall/blanket module because this structure experiences the greatest neutron fluence level and is the component for which the low swelling ferromagnetic Sandvik alloy, HT-9, may have the greatest benefit. The magnitudes of the magnetic body forces calculated were consistent with analyses performed on structures within other types of reactors. The loads generated within the module structure by the magnetic forces were found to be of the same order of magnitude as those arising from other sources such as pressure differential, dead weight, temperature distribution. Only small structural design modifications would be required if the magnetic alloy, Sandvik HT-9 were utilized.

Overview of the Lockheed Martin Compact Fusion Reactor (CFR) Project

NASA Astrophysics Data System (ADS)

McGuire, Thomas

2017-10-01

The Lockheed Martin Compact Fusion Reactor (CFR) Program endeavors to quickly develop a compact fusion power plant with favorable commercial economics and military utility. The CFR uses a diamagnetic, high beta, magnetically encapsulated, linear ring cusp plasma confinement scheme. Major project activities will be reviewed, including the T4B and T5 plasma heating experiments. The goal of the experiments is to demonstrate a suitable plasma target for heating experiments, to characterize the behavior of plasma sources in the CFR configuration and to then heat the plasma with neutral beams, with the plasma transitioning into the high Beta confinement regime. The design and preliminary results of the experiments will be presented, including discussion of predicted behavior, plasma sources, heating mechanisms, diagnostics suite and relevant numerical modeling. ©2017 Lockheed Martin Corporation. All Rights Reserved.

Alfvén oscillations in ohmic discharges with runaway electrons in the TUMAN-3M tokamak

NASA Astrophysics Data System (ADS)

Tukachinsky, A. S.; Askinazi, L. G.; Balachenkov, I. M.; Belokurov, A. A.; Gin, D. B.; Zhubr, N. A.; Kornev, V. A.; Lebedev, S. V.; Khil'kevich, E. M.; Chugunov, I. N.; Shevelev, A. E.

2016-12-01

Studying the mechanism of Alfvén wave generation in plasma is important, since the interaction of these waves with energetic particles in tokamak-type reactors can increase the losses of energy and particles with the corresponding decrease in the efficiency of plasma heating and, under certain conditions, lead to the damage of structural elements of the system. Despite the previous detailed investigations of the excitation of Alfvén waves by superthermal particles in regimes with additional heating, the physics of Alfvén mode generation in discharges with ohmic heating of plasma is still not sufficiently studied. We have established that a significant factor inf luencing the development of Alfvén oscillations in ohmic discharge is the presence of runaway electrons. A physical mechanism explaining this relationship is proposed.

Fully non-inductive second harmonic electron cyclotron plasma ramp-up in the QUEST spherical tokamak

NASA Astrophysics Data System (ADS)

Idei, H.; Kariya, T.; Imai, T.; Mishra, K.; Onchi, T.; Watanabe, O.; Zushi, H.; Hanada, K.; Qian, J.; Ejiri, A.; Alam, M. M.; Nakamura, K.; Fujisawa, A.; Nagashima, Y.; Hasegawa, M.; Matsuoka, K.; Fukuyama, A.; Kubo, S.; Shimozuma, T.; Yoshikawa, M.; Sakamoto, M.; Kawasaki, S.; Nakashima, H.; Higashijima, A.; Ide, S.; Maekawa, T.; Takase, Y.; Toi, K.

2017-12-01

Fully non-inductive second (2nd) harmonic electron cyclotron (EC) plasma current ramp-up was demonstrated with a newlly developed 28 GHz system in the QUEST spherical tokamak. A high plasma current of 54 kA was non-inductively ramped up and sustained stably for 0.9 s with a 270 kW 28 GHz wave. A higher plasma current of 66 kA was also non-inductively achieved with a slow ramp-up of the vertical field. We have achieved a significantly higher plasma current than those achieved previously with the 2nd harmonic EC waves. This fully non-inductive 2nd harmonic EC plasma ramp-up method might be useful for future burning plasma devices and fusion reactors, in particular for operations at half magnetic field with the same EC heating equipment.

The software-defined fast post-processing for GEM soft x-ray diagnostics in the Tungsten Environment in Steady-state Tokamak thermal fusion reactor

NASA Astrophysics Data System (ADS)

Krawczyk, Rafał Dominik; Czarski, Tomasz; Linczuk, Paweł; Wojeński, Andrzej; Kolasiński, Piotr; GÄ ska, Michał; Chernyshova, Maryna; Mazon, Didier; Jardin, Axel; Malard, Philippe; Poźniak, Krzysztof; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Kowalska-Strzeciwilk, Ewa; Malinowski, Karol

2018-06-01

This article presents a novel software-defined server-based solutions that were introduced in the fast, real-time computation systems for soft X-ray diagnostics for the WEST (Tungsten Environment in Steady-state Tokamak) reactor in Cadarache, France. The objective of the research was to provide a fast processing of data at high throughput and with low latencies for investigating the interplay between the particle transport and magnetohydrodynamic activity. The long-term objective is to implement in the future a fast feedback signal in the reactor control mechanisms to sustain the fusion reaction. The implemented electronic measurement device is anticipated to be deployed in the WEST. A standalone software-defined computation engine was designed to handle data collected at high rates in the server back-end of the system. Signals are obtained from the front-end field-programmable gate array mezzanine cards that acquire and perform a selection from the gas electron multiplier detector. A fast, authorial library for plasma diagnostics was written in C++. It originated from reference offline MATLAB implementations. They were redesigned for runtime analysis during the experiment in the novel online modes of operation. The implementation allowed the benchmarking, evaluation, and optimization of plasma processing algorithms with the possibility to check the consistency with reference computations written in MATLAB. The back-end software and hardware architecture are presented with data evaluation mechanisms. The online modes of operation for the WEST are discussed. The results concerning the performance of the processing and the introduced functionality are presented.

Design of a tokamak fusion reactor first wall armor against neutral beam impingement

DOE Office of Scientific and Technical Information (OSTI.GOV)

Myers, R.A.

1977-12-01

The maximum temperatures and thermal stresses are calculated for various first wall design proposals, using both analytical solutions and the TRUMP and SAP IV Computer Codes. Beam parameters, such as pulse time, cycle time, and beam power, are varied. It is found that uncooled plates should be adequate for near-term devices, while cooled protection will be necessary for fusion power reactors. Graphite and tungsten are selected for analysis because of their desirable characteristics. Graphite allows for higher heat fluxes compared to tungsten for similar pulse times. Anticipated erosion (due to surface effects) and plasma impurity fraction are estimated. Neutron irradiationmore » damage is also discussed. Neutron irradiation damage (rather than erosion, fatigue, or creep) is estimated to be the lifetime-limiting factor on the lifetime of the component in fusion power reactors. It is found that the use of tungsten in fusion power reactors, when directly exposed to the plasma, will cause serious plasma impurity problems; graphite should not present such an impurity problem.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bentz, L.K.; Bender, D.S.

This report contains socioeconomic information on the Plainsboro, New Jersey, area, the proposed location of the Compact Ignition Tokamak (CIT) facility. It was prepared as supplemental information for an environmental assessment for the CIT at Princeton Plasma Physics Laboratory (PPPL). The report contains descriptions of the demographic, economic, and community resource characteristics, and, based on information available in early 1987, considers the socioeconomic effect of the proposed facility. In all areas examined, the anticipated socioeconomic impacts of the proposed CIT facility at PPPL are negligible or minimal. 29 refs., 8 figs., 24 tabs.

The possible role of Reynolds stress in the creation of a transport barrier in tokamak edge plasmas

NASA Astrophysics Data System (ADS)

Vergote, M.; van Schoor, M.; Xu, Y.; Jachmich, S.; Weynants, R.; Hron, M.; Stöckel, J.

2005-03-01

To obtain a good confinement, mandatory in a fusion reactor, the understanding of the formation of transport barriers in the edge plasma of a tokamak is essential. Turbulence, the major candidate to explain anomalous transport, can be quenched by sheared flows in the edge which rip the convective cells apart, thus forming a barrier. Experimental evidence from the Chinese HT-6M tokamak [Y.H. Xu et al.: Phys. Rev. Lett. 84 (2000) 3867], points to the fact that momentum transfer from the turbulence can create these sheared flows via the Reynolds stresses. A new 1-D fluid model for the generation of the poloidal flow, has been developed taking into account the driving force of the Reynolds stress and the friction forces due to neutrals and parallel viscosity. Special attention has been dedicated to the computation of the flux-surface-averaging for the various terms. This model has been confronted with the experimental results obtained in the HT-6M tokamak, where Reynolds stresses were generated by application of a turbulent heating pulse. If the model is applied in cylindrical geometry, the calculated Reynolds stress-induced flow agrees well with the measured poloidal velocity in the plasma edge. However, when the full toroidal geometry is taken into account, it seems that the Reynolds stresses are too small to explain the observed rotation. This indicates that the role of the Reynolds stresses in inducing macroscopic flow in the torus is weakened. A combined system of probes allowing to measure the Reynolds stress and the rotation velocity simultaneously, has been developed and installed on the CASTOR tokamak (Prague). We report here on the first results obtained.

Metallic mirrors for plasma diagnosis in current and future reactors: tests for ITER and DEMO

NASA Astrophysics Data System (ADS)

Rubel, M.; Moon, Soonwoo; Petersson, P.; Garcia-Carrasco, A.; Hallén, A.; Krawczynska, A.; Fortuna-Zaleśna, E.; Gilbert, M.; Płociński, T.; Widdowson, A.; Contributors, JET

2017-12-01

Optical spectroscopy and imaging diagnostics in next-step fusion devices will rely on metallic mirrors. The performance of mirrors is studied in present-day tokamaks and in laboratory systems. This work deals with comprehensive tests of mirrors: (a) exposed in JET with the ITER-like wall (JET-ILW); (b) irradiated by hydrogen, helium and heavy ions to simulate transmutation effects and damage which may be induced by neutrons under reactor conditions. The emphasis has been on surface modification: deposited layers on JET mirrors from the divertor and on near-surface damage in ion-irradiated targets. Analyses performed with ion beams, microscopy and spectro-photometry techniques have revealed: (i) the formation of multiple co-deposited layers; (ii) flaking-off of the layers already in the tokamak, despite the small thickness (130-200 nm) of the granular deposits; (iii) deposition of dust particles (0.2-5 μm, 300-400 mm-2) composed mainly of tungsten and nickel; (iv) that the stepwise irradiation of up to 30 dpa by heavy ions (Mo, Zr or Nb) caused only small changes in the optical performance, in some cases even improving reflectivity due to the removal of the surface oxide layer; (v) significant reflectivity degradation related to bubble formation caused by the irradiation with He and H ions.

The LTX- β Research Program

NASA Astrophysics Data System (ADS)

Majeski, R.; Bell, R. E.; Boyle, D. P.; Hughes, P. E.; Kaita, R.; Kozub, T.; Merino, E.; Zhang, X.; Biewer, T. M.; Canik, J. M.; Elliott, D. B.; Reinke, M. L.; Bialek, J.; Hansen, C.; Jarboe, T.; Kubota, S.; Rhodes, T.; Dorf, M. A.; Rognlien, T.; Scotti, F.; Soukhanovskii, V. A.; Koel, B. E.; Donovan, D.; Maan, A.

2017-10-01

LTX- β, the upgrade to the Lithium Tokamak Experiment, approximately doubles the toroidal field (to 3.4 kG) and plasma current (to 150 - 175 kA) of LTX. Neutral beam injection at 20 kV, 30 A will be added in February 2018, with systems provided by Tri-Alpha Energy. A 9.3 GHz, 100 kW, short-pulse (5-10 msec) source will be available in summer 2018 for electron Bernstein wave heating. New lithium evaporation sources will allow between-shots recoating of the walls. Upgrades to the diagnostic set are intended to strengthen the research program in the critical areas of equilibrium, core transport, scrape-off layer physics, and plasma-material interactions. The LTX- β research program will combine the capability for gradient-free temperature profiles, to stabilize ion and electron temperature gradient-driven modes, with approaches to stabilization of ∇n-driven modes, such as the trapped electron mode (TEM). Candidate stabilization mechanisms for the TEM include sheared flow stabilization, which can be tested on LTX- β. The goal will be to minimize anomalous transport in a low aspect ratio tokamak, which would lead to a very compact, tokamak-based fusion core. This work supported by US DOE contracts DE-AC02-09CH11466 and DE-AC05-00OR22725.

Study of plasma-facing components in the Lithium Tokamak Experiment with the Materials Analysis and Particle Probe

NASA Astrophysics Data System (ADS)

Lucia, M.; Kaita, R.; Majeski, R.; Boyle, D. P.; Granstedt, E. M.; Jacobson, C. M.; Schmitt, J. C.; Allain, J. P.; Bedoya, F.; Gonderman, S.

2013-10-01

The Lithium Tokamak Experiment (LTX) is a spherical torus designed to accommodate solid or liquid lithium as the primary plasma-facing component (PFC). We present initial results from the implementation on LTX of the Materials Analysis and Particle Probe (MAPP) diagnostic, a collaboration among PPPL, Purdue University, and the University of Illinois. MAPP is a compact in vacuo surface science diagnostic, and its operation on LTX will provide the first ever in situ surface measurements of a tokamak first wall environment. With MAPP's analysis techniques, we will study the evolution of the surface chemistry of LTX's first wall as a function of varied temperature and lithium coating. During its 2013 run campaign, LTX will use an electron beam to evaporate lithium onto the first wall from an in-vessel reservoir. We will use two quartz crystal microbalances to estimate thickness of lithium coatings thus applied to the MAPP probe. We have recently installed a set of triple Langmuir probes on LTX, and they will be used to relate LTX edge plasma parameters to MAPP results. We will combine data from MAPP and the triple probes to estimate the local edge recycling coefficient based on desorption of retained hydrogen. This work was supported by U.S. DOE contract DE-AC02-09CH11466.

Studies of breakeven prices and electricity supply potentials of nuclear fusion by a long-term world energy and environment model

NASA Astrophysics Data System (ADS)

Tokimatsu, K.; Asaoka, Y.; Konishi, S.; Fujino, J.; Ogawa, Y.; Okano, K.; Nishio, S.; Yoshida, T.; Hiwatari, R.; Yamaji, K.

2002-11-01

In response to social demand, this paper investigates the breakeven price (BP) and potential electricity supply of nuclear fusion energy in the 21st century by means of a world energy and environment model. We set the following objectives in this paper: (i) to reveal the economics of the introduction conditions of nuclear fusion; (ii) to know when tokamak-type nuclear fusion reactors are expected to be introduced cost-effectively into future energy systems; (iii) to estimate the share in 2100 of electricity produced by the presently designed reactors that could be economically selected in the year. The model can give in detail the energy and environment technologies and price-induced energy saving, and can illustrate optimal energy supply structures by minimizing the costs of total discounted energy systems at a discount rate of 5%. The following parameters of nuclear fusion were considered: cost of electricity (COE) in the nuclear fusion introduction year, annual COE reduction rates, regional introduction year, and regional nuclear fusion capacity projection. The investigations are carried out for three nuclear fusion projections one of which includes tritium breeding constraints, four future CO2 concentration constraints, and technological assumptions on fossil fuels, nuclear fission, CO2 sequestration, and anonymous innovative technologies. It is concluded that: (1) the BPs are from 65 to 125 mill kW-1 h-1 depending on the introduction year of nuclear fusion under the 550 ppmv CO2 concentration constraints; those of a business-as-usual (BAU) case are from 51 to 68 mill kW-1h-1. Uncertainties resulting from the CO2 concentration constraints and the technological options influenced the BPs by plus/minus some 10 30 mill kW-1h-1, (2) tokamak-type nuclear fusion reactors (as presently designed, with a COE range around 70 130 mill kW-1h-1) would be favourably introduced into energy systems after 2060 based on the economic criteria under the 450 and 550 ppmv CO2 concentration constraint, but not selected under the BAU case and 650 ppmv CO2 concentration constraint, and (3) the share of electricity in 2100 produced by the presently designed tokamak-type nuclear fusion reactors (introduced after 2060) is well below 30%. It should be noted that these conclusions are based upon varieties of uncertainties in scenarios and data assumptions on nuclear fusion as well as technological options.

Progress in the RAMI analysis of a conceptual LHCD system for DEMO

NASA Astrophysics Data System (ADS)

Mirizzi, F.

2014-02-01

Reliability, Availability, Maintainability and Inspectability (RAMI) concepts and techniques, that acquired great importance during the first manned space missions, have been progressively extended to industrial, scientific and consumer equipments to assure them satisfactory performances and lifetimes. In the design of experimental facilities, like tokamaks, mainly aimed at demonstrating validity and feasibility of scientific theories, RAMI analysis has been often left aside. DEMO, the future prototype fusion reactors, will be instead designed for steadily delivering electrical energy to commercial grids, so that the RAMI aspects will assume an absolute relevance since their initial design phases. A preliminary RAMI analysis of the LHCD system for the conceptual EU DEMO reactor is given in the paper.

Bootstrap and fast wave current drive for tokamak reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ehst, D.A.

1991-09-01

Using the multi-species neoclassical treatment of Hirshman and Sigmar we study steady state bootstrap equilibria with seed currents provided by low frequency (ICRF) fast waves and with additional surface current density driven by lower hybrid waves. This study applies to reactor plasmas of arbitrary aspect ratio. IN one limit the bootstrap component can supply nearly the total equilibrium current with minimal driving power (< 20 MW). However, for larger total currents considerable driving power is required (for ITER: I{sub o} = 18 MA needs P{sub FW} = 15 MW, P{sub LH} = 75 MW). A computational survey of bootstrap fractionmore » and current drive efficiency is presented. 11 refs., 8 figs.« less

Articulated limiter blade for a tokamak fusion reactor

DOEpatents

Doll, D.W.

1982-10-21

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Articulated limiter blade for a tokamak fusion reactor

DOEpatents

Doll, David W.

1985-01-01

A limiter blade for a large tokomak fusion reactor includes three articulated blade sections for enabling the limiter blade to be adjusted for plasmas of different sizes. Each blade section is formed of a rigid backing plate carrying graphite tiles coated with titanium carbide, and the limiter blade forms a generally elliptic contour in both the poloidal and toroidal directions to uniformly distribute the heat flow to the blade. The limiter blade includes a central blade section movable along the major radius of the vacuum vessel, and upper and lower pivotal blade sections which may be pivoted by linear actuators having rollers held to the back surface of the pivotal blade sections.

Simultaneous C and N removal from saline salmon effluents in filter reactors comprising anoxic-anaerobic-aerobic processes: effect of recycle ratio.

PubMed

Giustinianovich, Elisa A; Aspé, Estrella R; Huiliñir, César E; Roeckel, Marlene D

2014-01-01

Salmon processing generates saline effluents with high protein load. To treat these effluents, three compact tubular filter reactors were installed and an integrated anoxic/anaerobic/aerobic process was developed with recycling flow from the reactor's exit to the inlet stream in order to save organic matter (OM) for denitrification. The reactors were aerated in the upper section with recycle ratios (RR) of 0, 2, and 10, respectively, at 30°C. A tubular reactor behave as a plug flow reactor when RR = 0, and as a mixed flow reactor when recycle increases, thus, different RR values were used to evaluate how it affects the product distribution and the global performance. Diluted salmon process effluent was prepared as substrate. Using loads of 1.0 kg COD m(-3)d(-1) and 0.15 kg total Kjeldahl nitrogen (TKN) m(-3)d(-1) at HRT of 2 d, 100% removal efficiencies for nitrite and nitrate were achieved in the anoxic-denitrifying section without effect of the dissolved oxygen in the recycled flow on denitrification. Removals >98% for total organic carbon (TOC) was achieved in the three reactors. The RR had no effect on the TOC removal; nevertheless a higher efficiency in total nitrogen removal in the reactor with the highest recycle ratio was observed: 94.3% for RR = 10 and 46.6% for RR = 2. Results showed that the proposed layout with an alternative distribution in a compact reactor can efficiently treat high organic carbon and nitrogen concentrations from a saline fish effluent with OM savings in denitrification.

Current Drive

DOE Office of Scientific and Technical Information (OSTI.GOV)

Faulconer, D.W

2004-03-15

Certain devices aimed at magnetic confinement of thermonuclear plasma rely on the steady flow of an electric current in the plasma. In view of the dominant place it occupies in both the world magnetic-confinement fusion effort and the author's own activity, the tokamak toroidal configuration is selected as prototype for discussing the question of how such a current can be maintained. Tokamaks require a stationary toroidal plasma current, this being traditionally provided by a pulsed magnetic induction which drives the plasma ring as the secondary of a transformer. Since this mechanism is essentially transient, and steady-state fusion reactor operation hasmore » manifold advantages, significant effort is now devoted to developing alternate steady-state means of generating toroidal current. These methods are classed under the global heading of 'noninductive current drive' or simply 'current drive', generally, though not exclusively, employing the injection of waves and/or toroidally directed particle beams. In what follows we highlight the physical mechanisms underlying surprisingly various approaches to driving current in a tokamak, downplaying a number of practical and technical issues. When a significant data base exists for a given method, its experimental current drive efficiency and future prospects are detailed.« less

Fully non-inductive plasma start-up with lower-hybrid waves using the outboard-launch and top-launch antennas on the TST-2 spherical tokamak

NASA Astrophysics Data System (ADS)

Tsujii, Naoto; Takase, Yuichi; Ejiri, Akira; Shinya, Takahiro; Yajima, Satoru; Yamazaki, Hibiki; Togashi, Hiro; Moeller, Charles P.; Roidl, Benedikt; Takahashi, Wataru; Toida, Kazuya; Yoshida, Yusuke

2017-10-01

Removal of the central solenoid is essential to realize an economical spherical tokamak fusion reactor, but non-inductive plasma start-up is a challenge. On the TST-2 spherical tokamak, non-inductive plasma start-up using lower-hybrid (LH) waves has been investigated. Using the capacitively-coupled combline (CCC) antenna installed at the outboard midplane, fully non-inductive plasma current ramp-up up to a quarter of that of the typical Ohmic discharges has been achieved. Although it was desirable to keep the density low during the plasma current ramp-up to avoid the LH density limit, it was recognized that there was a maximum current density that could be carried by a given electron density. Since the density needed to increase as the plasma current was ramped-up, the achievable plasma current was limited by the maximum operational toroidal field of TST-2. The top-launch CCC antenna was installed to access higher density with up-shift of the parallel index of refraction. Numerical analysis of LH current drive with the outboard-launch and top-launch antennas was performed and the results were qualitatively consistent with the experimental observations.

FABRICATION OF TUBE TYPE FUEL ELEMENT FOR NUCLEAR REACTORS

DOEpatents

Loeb, E.; Nicklas, J.H.

1959-02-01

A method of fabricating a nuclear reactor fuel element is given. It consists essentially of fixing two tubes in concentric relationship with respect to one another to provide an annulus therebetween, filling the annulus with a fissionablematerial-containing powder, compacting the powder material within the annulus and closing the ends thereof. The powder material is further compacted by swaging the inner surface of the inner tube to increase its diameter while maintaining the original size of the outer tube. This process results in reduced fabrication costs of powdered fissionable material type fuel elements and a substantial reduction in the peak core temperatures while materially enhancing the heat removal characteristics.

Comparison between instrumented precracked Charpy and compact specimen tests of carbon steels

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nanstad, R.K.

1980-01-01

The General Atomic Company High Temperature Gas-Cooled Reactor (HTGR) is housed within a prestressed concrete reactor vessel (PCRV). Various carbon steel structural members serve as closures at penetrations in the vessel. A program of testing and evaluation is underway to determine the need for reference fracture toughness (K/sub IR/) and indexing procedures for these materials as described in Appendix G to Section III, ASME Code for light water reactor steels. The materials of interest are carbon steel forgings (SA508, Class 1) and plates (SA537, Classes 1 and 2) as well as weldments of these steels. The fracture toughness behavior ismore » characterized with instrumented precracked Charpy V-votch specimens (PCVN) - slow-bend and dynamic - and compact specimens (10-mm and 25-mm thicknesses) using both linear elastic (ASTM E399) and elastic-plastic (equivalent Energy and J-Integral) analytical procedures. For the dynamic PCVN tests, force-time traces are analyzed according to the procedures of the Pressure Vessel Research Council (PVRC)/Metal Properties Council (MPC). Testing and analytical procedures are discussed and PCVN results are compared to those obtained with compact specimens.« less

REACTOR

DOEpatents

Christy, R.F.

1961-07-25

A means is described for co-relating the essential physical requirements of a fission chain reaction in order that practical, compact, and easily controllable reactors can be built. These objects are obtained by employing a composition of fissionsble isotope and moderator in fluid form in which the amount of fissionsble isotcpe present governs the reaction. The size of the reactor is no longer a critical factor, the new criterion being the concentration of the fissionable isotope.

A Compact Torus Fusion Reactor Utilizing a Continuously Generated Strings of CT's. The CT String Reactor, CTSR.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hartman, C W; Reisman, D B; McLean, H S

2007-05-30

A fusion reactor is described in which a moving string of mutually repelling compact toruses (alternating helicity, unidirectional Btheta) is generated by repetitive injection using a magnetized coaxial gun driven by continuous gun current with alternating poloidal field. An injected CT relaxes to a minimum magnetic energy equilibrium, moves into a compression cone, and enters a conducting cylinder where the plasma is heated to fusion-producing temperature. The CT then passes into a blanketed region where fusion energy is produced and, on emergence from the fusion region, the CT undergoes controlled expansion in an exit cone where an alternating poloidal fieldmore » opens the flux surfaces to directly recover the CT magnetic energy as current which is returned to the formation gun. The CT String Reactor (CTSTR) reactor satisfies all the necessary MHD stability requirements and is based on extrapolation of experimentally achieved formation, stability, and plasma confinement. It is supported by extensive 2D, MHD calculations. CTSTR employs minimal external fields supplied by normal conductors, and can produce high fusion power density with uniform wall loading. The geometric simplicity of CTSTR acts to minimize initial and maintenance costs, including periodic replacement of the reactor first wall.« less

Low-cost, compact, cooled photomultiplier assembly for use in magnetic fields up to 1400 Gauss

NASA Technical Reports Server (NTRS)

Patch, R. W.; Tashjian, R. A.; Jentner, T. A.

1975-01-01

Use of vortex tube for cooling and concentric shielding have produced smaller and more compact unit than was previously available. Future uses of device could include installation in gas chromatographs and mass spectrometers. Additional uses would include measurements and controls in magnetohydrodynamic power generators and fusion reactors.

Formation of the internal transport barrier in KSTAR

NASA Astrophysics Data System (ADS)

Chung, J.; Kim, H. S.; Jeon, Y. M.; Kim, J.; Choi, M. J.; Ko, J.; Lee, K. D.; Lee, H. H.; Yi, S.; Kwon, J. M.; Hahn, S.-H.; Ko, W. H.; Lee, J. H.; Yoon, S. W.

2018-01-01

One of key objectives of tokamak experiments is the exploration of enhanced confinement regimes, and the access of the internal transport barrier (ITB) formation is dealt with an important physics issue in the most of major tokamaks. Also, the advanced tokamak scenario with ITB is expected to lead to a continuous reactor with high fusion power density. From that point of view, the formation of the ITB in KSTAR which is designed for long pulse operation capability is very important although its heating and current drive systems are not fully equipped yet. We have therefore assumed that an early injection of the full NBI power (∼5.5 MW) during the current ramp-up would give a chance to form an internal barrier if the plasma could stay in the L-mode. To avoid the H-mode transition, we have produced inboard limited plasmas with detaching from the both upper and lower divertors. Using this approach, an ITB formation during L-mode has been observed which shows improved core confinement. Ion and electron temperature profiles show the barrier clearly in the temperature, and it was sustained for about 7 s in the dedicated experiment. This is the first stationary ITB observed in a full superconducting tokamak. This operation scenario with the ITB could be an alternative way to achieve a high performance regime in KSTAR, and the length of the ITB discharge could be extended even longer. In this paper, we present the formation of the ITB using measured and simulated characteristic profiles.

Overview of International Thermonuclear Experimental Reactor (ITER) engineering design activities*

NASA Astrophysics Data System (ADS)

Shimomura, Y.

1994-05-01

The International Thermonuclear Experimental Reactor (ITER) [International Thermonuclear Experimental Reactor (ITER) (International Atomic Energy Agency, Vienna, 1988), ITER Documentation Series, No. 1] project is a multiphased project, presently proceeding under the auspices of the International Atomic Energy Agency according to the terms of a four-party agreement among the European Atomic Energy Community (EC), the Government of Japan (JA), the Government of the Russian Federation (RF), and the Government of the United States (US), ``the Parties.'' The ITER project is based on the tokamak, a Russian invention, and has since been brought to a high level of development in all major fusion programs in the world. The objective of ITER is to demonstrate the scientific and technological feasibility of fusion energy for peaceful purposes. The ITER design is being developed, with support from the Parties' four Home Teams and is in progress by the Joint Central Team. An overview of ITER Design activities is presented.

New design of cable-in-conduit conductor for application in future fusion reactors

NASA Astrophysics Data System (ADS)

Qin, Jinggang; Wu, Yu; Li, Jiangang; Liu, Fang; Dai, Chao; Shi, Yi; Liu, Huajun; Mao, Zhehua; Nijhuis, Arend; Zhou, Chao; Yagotintsev, Konstantin A.; Lubkemann, Ruben; Anvar, V. A.; Devred, Arnaud

2017-11-01

The China Fusion Engineering Test Reactor (CFETR) is a new tokamak device whose magnet system includes toroidal field, central solenoid (CS) and poloidal field coils. The main goal is to build a fusion engineering tokamak reactor with about 1 GW fusion power and self-sufficiency by blanket. In order to reach this high performance, the magnet field target is 15 T. However, the huge electromagnetic load caused by high field and current is a threat for conductor degradation under cycling. The conductor with a short-twist-pitch (STP) design has large stiffness, which enables a significant performance improvement in view of load and thermal cycling. But the conductor with STP design has a remarkable disadvantage: it can easily cause severe strand indentation during cabling. The indentation can reduce the strand performance, especially under high load cycling. In order to overcome this disadvantage, a new design is proposed. The main characteristic of this new design is an updated layout in the triplet. The triplet is made of two Nb3Sn strands and one soft copper strand. The twist pitch of the two Nb3Sn strands is large and cabled first. The copper strand is then wound around the two superconducting strands (CWS) with a shorter twist pitch. The following cable stages layout and twist pitches are similar to the ITER CS conductor with STP design. One short conductor sample with a similar scale to the ITER CS was manufactured and tested with the Twente Cable Press to investigate the mechanical properties, AC loss and internal inspection by destructive examination. The results are compared to the STP conductor (ITER CS and CFETR CSMC) tests. The results show that the new conductor design has similar stiffness, but much lower strand indentation than the STP design. The new design shows potential for application in future fusion reactors.

Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kane, J. J.; van Rooyen, I. J.; Craft, A. E.

In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

Evaluating quantitative 3-D image analysis as a design tool for low enriched uranium fuel compacts for the transient reactor test facility: A preliminary study

DOE PAGES

Kane, J. J.; van Rooyen, I. J.; Craft, A. E.; ...

2016-02-05

In this study, 3-D image analysis when combined with a non-destructive examination technique such as X-ray computed tomography (CT) provides a highly quantitative tool for the investigation of a material’s structure. In this investigation 3-D image analysis and X-ray CT were combined to analyze the microstructure of a preliminary subsized fuel compact for the Transient Reactor Test Facility’s low enriched uranium conversion program to assess the feasibility of the combined techniques for use in the optimization of the fuel compact fabrication process. The quantitative image analysis focused on determining the size and spatial distribution of the surrogate fuel particles andmore » the size, shape, and orientation of voids within the compact. Additionally, the maximum effect of microstructural features on heat transfer through the carbonaceous matrix of the preliminary compact was estimated. The surrogate fuel particles occupied 0.8% of the compact by volume with a log-normal distribution of particle sizes with a mean diameter of 39 μm and a standard deviation of 16 μm. Roughly 39% of the particles had a diameter greater than the specified maximum particle size of 44 μm suggesting that the particles agglomerate during fabrication. The local volume fraction of particles also varies significantly within the compact although uniformities appear to be evenly dispersed throughout the analysed volume. The voids produced during fabrication were on average plate-like in nature with their major axis oriented perpendicular to the compaction direction of the compact. Finally, the microstructure, mainly the large preferentially oriented voids, may cause a small degree of anisotropy in the thermal diffusivity within the compact. α∥/α⊥, the ratio of thermal diffusivities parallel to and perpendicular to the compaction direction are expected to be no less than 0.95 with an upper bound of 1.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Stewart Zweben; Samuel Cohen; Hantao Ji

Small ''concept exploration'' experiments have for many years been an important part of the fusion research program at the Princeton Plasma Physics Laboratory (PPPL). this paper describes some of the present and planned fusion concept exploration experiments at PPPL. These experiments are a University-scale research level, in contrast with the larger fusion devices at PPPL such as the National Spherical Torus Experiment (NSTX) and the Tokamak Fusion Test Reactor (TFTR), which are at ''proof-of-principle'' and ''proof-of-performance'' levels, respectively.

Readiness Review of BWXT for Fabrication of AGR 5/6/7 Compacts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Marshall, Douglas William; Sharp, Michelle Tracy

In support of preparations for fabricating compacts for the Advanced Gas Reactor (AGR) fuel qualification irradiation experiments (AGR-5/6/7), Idaho National Laboratory (INL) conducted a readiness review of the BWX Technology (BWXT) procedures, processes, and equipment associated with compact fabrication activities at the BWXT Nuclear Operations Group (BWXT-NOG) facility outside Lynchburg, VirginiaVA. The readiness review used quality assurance requirements taken from the American Society of Mechanical Engineers (ASME) Nuclear Quality Assurance Standard (NQA-1-2008/1a-2009) as a basis to assess readiness to start compact fabrication.

Physics of collisionless scrape-off-layer plasma during normal and off-normal Tokamak operating conditions.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hassanein, A.; Konkashbaev, I.

1999-03-15

The structure of a collisionless scrape-off-layer (SOL) plasma in tokamak reactors is being studied to define the electron distribution function and the corresponding sheath potential between the divertor plate and the edge plasma. The collisionless model is shown to be valid during the thermal phase of a plasma disruption, as well as during the newly desired low-recycling normal phase of operation with low-density, high-temperature, edge plasma conditions. An analytical solution is developed by solving the Fokker-Planck equation for electron distribution and balance in the SOL. The solution is in good agreement with numerical studies using Monte-Carlo methods. The analytical solutionsmore » provide an insight to the role of different physical and geometrical processes in a collisionless SOL during disruptions and during the enhanced phase of normal operation over a wide range of parameters.« less

GEM detectors development for radiation environment: neutron tests and simulations

NASA Astrophysics Data System (ADS)

Chernyshova, Maryna; Jednoróg, Sławomir; Malinowski, Karol; Czarski, Tomasz; Ziółkowski, Adam; Bieńkowska, Barbara; Prokopowicz, Rafał; Łaszyńska, Ewa; Kowalska-Strzeciwilk, Ewa; Poźniak, Krzysztof T.; Kasprowicz, Grzegorz; Zabołotny, Wojciech; Wojeński, Andrzej; Krawczyk, Rafał D.; Linczuk, Paweł; Potrykus, Paweł; Bajdel, Barcel

2016-09-01

One of the requests from the ongoing ITER-Like Wall Project is to have diagnostics for Soft X-Ray (SXR) monitoring in tokamak. Such diagnostics should be focused on tungsten emission measurements, as an increased attention is currently paid to tungsten due to a fact that it became a main candidate for the plasma facing material in ITER and future fusion reactor. In addition, such diagnostics should be able to withstand harsh radiation environment at tokamak during its operation. The presented work is related to the development of such diagnostics based on Gas Electron Multiplier (GEM) technology. More specifically, an influence of neutron radiation on performance of the GEM detectors is studied both experimentally and through computer simulations. The neutron induced radioactivity (after neutron source exposure) was found to be not pronounced comparing to an impact of other secondary neutron reaction products (during the exposure).

Non-inductively driven tokamak plasmas at near-unity βt in the Pegasus toroidal experiment

NASA Astrophysics Data System (ADS)

Reusch, J. A.; Bodner, G. M.; Bongard, M. W.; Burke, M. G.; Fonck, R. J.; Pachicano, J. L.; Perry, J. M.; Pierren, C.; Rhodes, A. T.; Richner, N. J.; Rodriguez Sanchez, C.; Schlossberg, D. J.; Weberski, J. D.

2018-05-01

A major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓi, high elongation κ, and high toroidal and normalized beta ( βt and βN) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓi. The low aspect ratio ( R0/a ˜1.2 ) of Pegasus allows access to high κ and high normalized plasma currents ( IN=Ip/a BT>14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high βt plasmas. Equilibrium analyses indicate that βt up to ˜100% is achieved. These high βt discharges disrupt at the ideal no-wall β limit at βN˜7.

Modeling of Steady-state Scenarios for the Fusion Nuclear Science Facility, Advanced Tokamak Approach

NASA Astrophysics Data System (ADS)

Garofalo, A. M.; Chan, V. S.; Prater, R.; Smith, S. P.; St. John, H. E.; Meneghini, O.

2013-10-01

A Fusion National Science Facility (FNSF) would complement ITER in addressing the community identified science and technology gaps to a commercially attractive DEMO, including breeding tritium and completing the fuel cycle, qualifying nuclear materials for high fluence, developing suitable materials for the plasma-boundary interface, and demonstrating power extraction. Steady-state plasma operation is highly desirable to address the requirements for fusion nuclear technology testing [1]. The Advanced Tokamak (AT) is a strong candidate for an FNSF as a consequence of its mature physics base, capability to address the key issues with a more compact device, and the direct relevance to an attractive target power plant. Key features of AT are fully noninductive current drive, strong plasma cross section shaping, internal profiles consistent with high bootstrap fraction, and operation at high beta, typically above the free boundary limit, βN > 3 . Work supported by GA IR&D funding, DE-FC02-04ER54698, and DE-FG02-95ER43309.

NEET Micro-Pocket Fission Detector. Final Project report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Unruh, T.; Rempe, Joy; McGregor, Douglas

2014-09-01

A collaboration between the Idaho National Laboratory (INL), the Kansas State University (KSU), and the French Alternative Energies and Atomic Energy Commission, Commissariat à l'Énergie Atomique et aux Energies Alternatives, (CEA), is funded by the Nuclear Energy Enabling Technologies (NEET) program to develop and test Micro-Pocket Fission Detectors (MPFDs), which are compact fission chambers capable of simultaneously measuring thermal neutron flux, fast neutron flux and temperature within a single package. When deployed, these sensors will significantly advance flux detection capabilities for irradiation tests in US Material Test Reactors (MTRs). Ultimately, evaluations may lead to a more compact, more accurate, andmore » longer lifetime flux sensor for critical mock-ups, and high performance reactors, allowing several Department of Energy Office of Nuclear Energy (DOE-NE) programs to obtain higher accuracy/higher resolution data from irradiation tests of candidate new fuels and materials. Specifically, deployment of MPFDs will address several challenges faced in irradiations performed at MTRs: Current fission chamber technologies do not offer the ability to measure fast flux, thermal flux and temperature within a single compact probe; MPFDs offer this option. MPFD construction is very different than current fission chamber construction; the use of high temperature materials allow MPFDs to be specifically tailored to survive harsh conditions encountered in-core of high performance MTRs. The higher accuracy, high fidelity data available from the compact MPFD will significantly enhance efforts to validate new high-fidelity reactor physics codes and new multi-scale, multi-physics codes. MPFDs can be built with variable sensitivities to survive the lifetime of an experiment or fuel assembly in some MTRs, allowing for more efficient and cost effective power monitoring. The small size of the MPFDs allows multiple sensors to be deployed, offering the potential to accurately measure the flux and temperature profiles in the reactor. This report summarizes the status at the end of year two of this three year project. As documented in this report, all planned accomplishments for developing this unique new, compact, multipurpose sensor have been completed.« less

Development of plasma bolometers using fiber-optic temperature sensors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reinke, M. L., E-mail: reinkeml@ornl.gov; Han, M.; Liu, G.

Measurements of radiated power in magnetically confined plasmas are important for exhaust studies in present experiments and expected to be a critical diagnostic for future fusion reactors. Resistive bolometer sensors have long been utilized in tokamaks and helical devices but suffer from electromagnetic interference (EMI). Results are shown from initial testing of a new bolometer concept based on fiber-optic temperature sensor technology. A small, 80 μm diameter, 200 μm long silicon pillar attached to the end of a single mode fiber-optic cable acts as a Fabry–Pérot cavity when broadband light, λ{sub o} ∼ 1550 nm, is transmitted along the fiber.more » Changes in temperature alter the optical path length of the cavity primarily through the thermo-optic effect, resulting in a shift of fringes reflected from the pillar detected using an I-MON 512 OEM spectrometer. While initially designed for use in liquids, this sensor has ideal properties for use as a plasma bolometer: a time constant, in air, of ∼150 ms, strong absorption in the spectral range of plasma emission, immunity to local EMI, and the ability to measure changes in temperature remotely. Its compact design offers unique opportunities for integration into the vacuum environment in places unsuitable for a resistive bolometer. Using a variable focus 5 mW, 405 nm, modulating laser, the signal to noise ratio versus power density of various bolometer technologies are directly compared, estimating the noise equivalent power density (NEPD). Present tests show the fiber-optic bolometer to have NEPD of 5-10 W/m{sup 2} when compared to those of the resistive bolometer which can achieve <0.5 W/m{sup 2} in the laboratory, but this can degrade to 1-2 W/m{sup 2} or worse when installed on a tokamak. Concepts are discussed to improve the signal to noise ratio of this new fiber-optic bolometer by reducing the pillar height and adding thin metallic coatings, along with improving the spectral resolution of the interrogator.« less

Low-power lead-cooled fast reactor loaded with MOX-fuel

NASA Astrophysics Data System (ADS)

Sitdikov, E. R.; Terekhova, A. M.

2017-01-01

Fast reactor for the purpose of implementation of research, education of undergraduate and doctoral students in handling innovative fast reactors and training specialists for atomic research centers and nuclear power plants (BRUTs) was considered. Hard neutron spectrum achieved in the fast reactor with compact core and lead coolant. Possibility of prompt neutron runaway of the reactor is excluded due to the low reactivity margin which is less than the effective fraction of delayed neutrons. The possibility of using MOX fuel in the BRUTs reactor was examined. The effect of Keff growth connected with replacement of natural lead coolant to 208Pb coolant was evaluated. The calculations and reactor core model were performed using the Serpent Monte Carlo code.

Pellets for fusion reactor refueling. Annual progress report, January 1, 1976--December 31, 1976

DOE Office of Scientific and Technical Information (OSTI.GOV)

Turnbull, R. J.; Kim, K.

1977-01-01

The purpose of this research is to test the feasibility of refueling fusion reactors using solid pellets composed of fuel elements. A solid hydrogen pellet generator has been constructed and experiments have been done to inject the pellets into the ORMAK Tokamak. A theory has been developed to describe the pellet ablation in the plasma, and an excellent agreement has been found between the theory and the experiment. Techniques for charging the pellets have been developed in order to accelerate and control them. Other works currently under way include the development of techniques for accelerating the pellets for refueling purpose.more » Evaluation of electrostatic acceleration has also been performed.« less

Fusion pumped laser

DOEpatents

Pappas, D.S.

1987-07-31

The apparatus of this invention may comprise a system for generating laser radiation from a high-energy neutron source. The neutron source is a tokamak fusion reactor generating a long pulse of high-energy neutrons and having a temperature and magnetic field effective to generate a neutron flux of at least 10/sup 15/ neutrons/cm/sup 2//center dot/s. Conversion means are provided adjacent the fusion reactor at a location operable for converting the high-energy neutrons to an energy source with an intensity and energy effective to excite a preselected lasing medium. A lasing medium is spaced about and responsive to the energy source to generate a population inversion effective to support laser oscillations for generating output radiation. 2 figs., 2 tabs.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Neilson, G. H.; Heitzenroeder, P.; Lyon, J.

Stellarators use 3D plasma and magnetic field shaping to produce a steady-state disruption-free magnetic confinement configuration. Compact stellarators have additional attractive properties — quasi-symmetric magnetic fields and low aspect ratio. The National Compact Stellarator Experiment (NCSX) is being constructed at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL) to test the physics of a high-beta compact stellarator with a lowripple, tokamak-like magnetic configuration. The engineering challenges of NCSX stem from its complex geometry requirements. These issues are addressed in the construction project through manufacturing R&D and system engineering. As a result, the fabricationmore » of the coil winding forms and vacuum vessel are proceeding in industry without significant technical issues, and preparations for winding the coils at PPPL are in place. Design integration, analysis, and dimensional control are functions provided by system engineering to ensure that the finished product will satisfy the physics requirements, especially accurate realization of the specified coil geometries. After completion of construction in 2009, a research program to test the expected physics benefits will start.« less

PROCESS OF FORMING POWDERED MATERIAL

DOEpatents

Glatter, J.; Schaner, B.E.

1961-07-14

A process of forming high-density compacts of a powdered ceramic material is described by agglomerating the powdered ceramic material with a heat- decompossble binder, adding a heat-decompossble lubricant to the agglomerated material, placing a quantity of the material into a die cavity, pressing the material to form a compact, pretreating the compacts in a nonoxidizing atmosphere to remove the binder and lubricant, and sintering the compacts. When this process is used for making nuclear reactor fuel elements, the ceramic material is an oxide powder of a fissionsble material and after forming, the compacts are placed in a cladding tube which is closed at its ends by vapor tight end caps, so that the sintered compacts are held in close contact with each other and with the interior wall of the cladding tube.

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason Michael; Stempien, John Dennis; Demkowicz, Paul Andrew

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO 2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. Thesemore » data were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO 2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO 2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

Fission Product Inventory and Burnup Evaluation of the AGR-2 Irradiation by Gamma Spectrometry

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

Gamma spectrometry has been used to evaluate the burnup and fission product inventory of different components from the US Advanced Gas Reactor Fuel Development and Qualification Program's second TRISO-coated particle fuel irradiation test (AGR-2). TRISO fuel in this irradiation included both uranium carbide / uranium oxide (UCO) kernels and uranium oxide (UO2) kernels. Four of the 6 capsules contained fuel from the US Advanced Gas Reactor program, and only those capsules will be discussed in this work. The inventories of gamma-emitting fission products from the fuel compacts, graphite compact holders, graphite spacers and test capsule shell were evaluated. These datamore » were used to measure the fractional release of fission products such as Cs-137, Cs-134, Eu-154, Ce-144, and Ag-110m from the compacts. The fraction of Ag-110m retained in the compacts ranged from 1.8% to full retention. Additionally, the activities of the radioactive cesium isotopes (Cs-134 and Cs-137) have been used to evaluate the burnup of all US TRISO fuel compacts in the irradiation. The experimental burnup evaluations compare favorably with burnups predicted from physics simulations. Predicted burnups for UCO compacts range from 7.26 to 13.15 % fission per initial metal atom (FIMA) and 9.01 to 10.69 % FIMA for UO2 compacts. Measured burnup ranged from 7.3 to 13.1 % FIMA for UCO compacts and 8.5 to 10.6 % FIMA for UO2 compacts. Results from gamma emission computed tomography performed on compacts and graphite holders that reveal the distribution of different fission products in a component will also be discussed. Gamma tomography of graphite holders was also used to locate the position of TRISO fuel particles suspected of having silicon carbide layer failures that lead to in-pile cesium release.« less

An in situ accelerator-based diagnostic for plasma-material interactions science on magnetic fusion devices.

PubMed

Hartwig, Zachary S; Barnard, Harold S; Lanza, Richard C; Sorbom, Brandon N; Stahle, Peter W; Whyte, Dennis G

2013-12-01

This paper presents a novel particle accelerator-based diagnostic that nondestructively measures the evolution of material surface compositions inside magnetic fusion devices. The diagnostic's purpose is to contribute to an integrated understanding of plasma-material interactions in magnetic fusion, which is severely hindered by a dearth of in situ material surface diagnosis. The diagnostic aims to remotely generate isotopic concentration maps on a plasma shot-to-shot timescale that cover a large fraction of the plasma-facing surface inside of a magnetic fusion device without the need for vacuum breaks or physical access to the material surfaces. Our instrument uses a compact (~1 m), high-current (~1 milliamp) radio-frequency quadrupole accelerator to inject 0.9 MeV deuterons into the Alcator C-Mod tokamak at MIT. We control the tokamak magnetic fields--in between plasma shots--to steer the deuterons to material surfaces where the deuterons cause high-Q nuclear reactions with low-Z isotopes ~5 μm into the material. The induced neutrons and gamma rays are measured with scintillation detectors; energy spectra analysis provides quantitative reconstruction of surface compositions. An overview of the diagnostic technique, known as accelerator-based in situ materials surveillance (AIMS), and the first AIMS diagnostic on the Alcator C-Mod tokamak is given. Experimental validation is shown to demonstrate that an optimized deuteron beam is injected into the tokamak, that low-Z isotopes such as deuterium and boron can be quantified on the material surfaces, and that magnetic steering provides access to different measurement locations. The first AIMS analysis, which measures the relative change in deuterium at a single surface location at the end of the Alcator C-Mod FY2012 plasma campaign, is also presented.

Vacuum system design and tritium inventory for the charge exchange diagnostic on the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Medley, S.S.

The application of charge exchange analyzers for the measurement of ion temperature in fusion plasma experiments requires a direct connection between the diagnostic and plasma-discharge vacuum chambers. Differential pumping of the gas load from the diagnostic stripping cell operated at > or approx. = 10/sup -3/ Torr is required to maintain the analyzer chamber at a pressure of < or approx. = 10/sup -6/ Torr. The migration of gases between the diagnostic and plasma vacuum chambers must be minimized. In particular, introduction of the analyzer stripping cell gas into the plasma chamber having a base pressure of < or approx.more » = 10/sup -8/ Torr must be suppressed. The charge exchange diagnostic for the Tokamak Fusion Test Reactor (TFTR) is comprised of two analyzer systems designed to contain a total of 18 independent mass/energy analyzers and one diagnostic neutral beam rated at 80 keV, 15 A. The associated arrays of multiple, interconnected vacuum systems were analyzed using the Vacuum System Transient Simulator (Vsts) computer program which models the transient transport of multigas species through complex networks of ducts, valves, traps, vacuum pumps, and other related vacuum system components. In addition to providing improved design performance at reduced costs, the analysis yields estimates for the exchange of tritium from the torus to the diagnostic components and of the diagnostic working gases to the torus.« less

Impact and effects of simultaneous MeV-ion irradiation and helium plasma exposure to the formation of tungsten nano-tendrils

NASA Astrophysics Data System (ADS)

Wright, Graham; Kesler, Leigh Ann; Whyte, Dennis

2013-10-01

The extrusion of nano-tendrils from high temperature (>1000 K) tungsten (W) targets exposed to helium (He) plasma ions remains a concern for future fusion reactors. Previous work on the Alcator C-Mod tokamak has demonstrated it is possible to form these structures in a tokamak environment. However, one area where Alcator C-Mod and a fusion reactor differ is total neutron flux at the wall and the displacement damage these neutrons produce in the plasma-facing materials. This dsiplacement damage may affect the size and number He bubbles precipitating in the W target, which is a key factor in the formation and growth of the nano-tendrils. The DIONISOS experiment directly measures the impact of the displacement damage by simultaneously bombarding high temperature W targets with MeV-range ions (to simulate the displacement damage caused by neutron flux) and high flux of He plasma ions. Different combinations of irradiating ion species and W target temperatures are used to vary the different processes and rates that are involved such as He trapping rate, vacancy production and annealing rates, and nano-tendril growth rate. The nano-tendril growth is characterized by SEM imaging and focused ion beam (FIB) cross-sectioning and compared to nano-tendril formation without the presence of the irradiating ion beam. This work is supported by US DOE award DE-SC00-02060.

NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0.8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

NASA Astrophysics Data System (ADS)

Dickens, J. K.; Hill, N. W.; Hou, F. S.; McConnell, J. W.; Spencer, R. R.; Tsang, F. Y.

1985-08-01

A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in the detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.

ADX: A high Power Density, Advanced RF-Driven Divertor Test Tokamak for PMI studies

NASA Astrophysics Data System (ADS)

Whyte, Dennis; ADX Team

2015-11-01

The MIT PSFC and collaborators are proposing an advanced divertor experiment, ADX; a divertor test tokamak dedicated to address critical gaps in plasma-material interactions (PMI) science, and the world fusion research program, on the pathway to FNSF/DEMO. Basic ADX design features are motivated and discussed. In order to assess the widest range of advanced divertor concepts, a large fraction (>50%) of the toroidal field volume is purpose-built with innovative magnetic topology control and flexibility for assessing different surfaces, including liquids. ADX features high B-field (>6 Tesla) and high global power density (P/S ~ 1.5 MW/m2) in order to access the full range of parallel heat flux and divertor plasma pressures foreseen for reactors, while simultaneously assessing the effect of highly dissipative divertors on core plasma/pedestal. Various options for efficiently achieving high field are being assessed including the use of Alcator technology (cryogenic cooled copper) and high-temperature superconductors. The experimental platform would also explore advanced lower hybrid current drive and ion-cyclotron range of frequency actuators located at the high-field side; a location which is predicted to greatly reduce the PMI effects on the launcher while minimally perturbing the core plasma. The synergistic effects of high-field launchers with high total B on current and flow drive can thus be studied in reactor-relevant boundary plasmas.

Helium Catalyzed D-D Fusion in a Levitated Dipole

NASA Astrophysics Data System (ADS)

Kesner, J.; Bromberg, L.; Garnier, D. T.; Hansen, A.; Mauel, M. E.

2003-10-01

Fusion research has focused on the goal of deuterium and tritium (D-T) fusion power because the reaction rate is large compared with the other fusion fuels: D-D or D-He3. Furthermore, the D-D cycle is difficult in traditional confinement devices, such as tokamaks, because good energy confinement is accompanied by good particle confinement which leads to an accumulation of ash. Fusion reactors based on the D-D reaction would be advantageous to D-T based reactors since they do not require the breeding of tritium and can reduce the flux of energetic neutrons that cause material damage. We propose a fusion power source based on the levitated dipole fusion concept that uses a "helium catalyzed D-D" fuel cycle, where rapid circulation of plasma allows the removal of tritium and the re-injection of the He3 decay product, eliminating the need for a massive blanket and shield. Stable dipole confinement derives from plasma compressibility instead of the magnetic shear and average good curvature. As a result, a dipole magnetic field can stabilize plasma at high beta while allowing large-scale adiabatic particle circulation. These properties may make the levitated dipole uniquely capable of achieving good energy confinement with low particle confinement. We find that a dipole based D-D power source can provide better utilization of magnetic field energy with a comparable mass power density to a D-T based tokamak power source.

A mechanism for beam-driven excitation of ion cyclotron harmonic waves in the Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dendy, R.O.; McClements, K.G.; Lashmore-Davies, C.N.

1994-10-01

A mechanism is proposed for the excitation of waves at harmonics of the injected ion cyclotron frequencies in neutral beam-heated discharges in the Tokamak Fusion Test Reactor (TFTR) [[ital Proceedings] [ital of] [ital the] 17[ital th] [ital European] [ital Conference] [ital on] [ital Controlled] [ital Fusion] [ital and] [ital Plasma] [ital Heating] (European Physical Society, Petit-Lancy, Switzerland, 1990), p. 1540]. Such waves are observed to originate from the outer midplane edge of the plasma. It is shown that ion cyclotron harmonic waves can be destabilized by a low concentration of sub-Alfvenic deuterium or tritium beam ions, provided these ions havemore » a narrow distribution of speeds parallel to the magnetic field. Such a distribution is likely to occur in the edge plasma, close to the point of beam injection. The predicted instability gives rise to wave emission at propagation angles lying almost perpendicular to the field. In contrast to the magnetoacoustic cyclotron instability proposed as an excitation mechanism for fusion-product-driven ion cyclotron emission in the Joint European Torus (JET) [Phys. Plasmas [bold 1], 1918 (1994)], the instability proposed here does not involve resonant fast Alfven and ion Bernstein waves, and can be driven by sub-Alfvenic energetic ions. It is concluded that the observed emission from TFTR can be driven by beam ions.« less

Design Features of the Neutral Particle Diagnostic System for the ITER Tokamak

NASA Astrophysics Data System (ADS)

Petrov, S. Ya.; Afanasyev, V. I.; Melnik, A. D.; Mironov, M. I.; Navolotsky, A. S.; Nesenevich, V. G.; Petrov, M. P.; Chernyshev, F. V.; Kedrov, I. V.; Kuzmin, E. G.; Lyublin, B. V.; Kozlovski, S. S.; Mokeev, A. N.

2017-12-01

The control of the deuterium-tritium (DT) fuel isotopic ratio has to ensure the best performance of the ITER thermonuclear fusion reactor. The diagnostic system described in this paper allows the measurement of this ratio analyzing the hydrogen isotope fluxes (performing neutral particle analysis (NPA)). The development and supply of the NPA diagnostics for ITER was delegated to the Russian Federation. The diagnostics is being developed at the Ioffe Institute. The system consists of two analyzers, viz., LENPA (Low Energy Neutral Particle Analyzer) with 10-200 keV energy range and HENPA (High Energy Neutral Particle Analyzer) with 0.1-4.0MeV energy range. Simultaneous operation of both analyzers in different energy ranges enables researchers to measure the DT fuel ratio both in the central burning plasma (thermonuclear burn zone) and at the edge as well. When developing the diagnostic complex, it was necessary to account for the impact of several factors: high levels of neutron and gamma radiation, the direct vacuum connection to the ITER vessel, implying high tritium containment, strict requirements on reliability of all units and mechanisms, and the limited space available for accommodation of the diagnostic hardware at the ITER tokamak. The paper describes the design of the diagnostic complex and the engineering solutions that make it possible to conduct measurements under tokamak reactor conditions. The proposed engineering solutions provide a safe—with respect to thermal and mechanical loads—common vacuum channel for hydrogen isotope atoms to pass to the analyzers; ensure efficient shielding of the analyzers from the ITER stray magnetic field (up to 1 kG); provide the remote control of the NPA diagnostic complex, in particular, connection/disconnection of the NPA vacuum beamline from the ITER vessel; meet the ITER radiation safety requirements; and ensure measurements of the fuel isotopic ratio under high levels of neutron and gamma radiation.

Catalyzed D-D stellarator reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sheffield, John; Spong, Donald A.

The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, F R = 0.9 to 1.15, ≈ 8.0% to 11.5%, Z eff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, B m ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

Catalyzed D-D stellarator reactor

DOE PAGES

Sheffield, John; Spong, Donald A.

2016-05-12

The advantages of using the catalyzed deuterium-deuterium (D-D) approach for a fusion reactor—lower and less energetic neutron flux and no need for a tritium breeding blanket—have been evaluated in previous papers, giving examples of both tokamak and stellarator reactors. This paper presents an update for the stellarator example, taking account of more recent empirical transport scaling results and design studies of lower-aspect-ratio stellarators. We use a modified version of the Generic Magnetic Fusion Reactor model to cost a stellarator-type reactor. Recently, this model has been updated to reflect the improved science and technology base and costs in the magnetic fusionmore » program. Furthermore, it is shown that an interesting catalyzed D-D, stellarator power plant might be possible if the following parameters could be achieved: R/ ≈ 4, required improvement factor to ISS04 scaling, F R = 0.9 to 1.15, ≈ 8.0% to 11.5%, Z eff ≈ 1.45 plus a relativistic temperature correction, fraction of fast ions lost ≈ 0.07, B m ≈ 14 to 16 T, and R ≈ 18 to 24 m.« less

Design Status of the Cryogenic System and Operation Modes Analysys of the JT-60SA Tokamak

NASA Astrophysics Data System (ADS)

Roussel, P.; Hoa, C.; Lamaison, V.; Michel, F.; Reynaud, P.; Wanner, M.

2010-04-01

The JT-60SA project is part of the Broader Approach Programme signed between Japan and Europe. This superconducting upgrade of the existing JT-60U tokamak in Naka, Japan shall start operation in 2016 and shall support ITER exploitation and research towards DEMO fusion reactor. JT-60SA is currently in the basic design phase. The cryogenic system of JT-60SA shall provide supercritical helium to cool the superconducting magnets and their structures at 4.4 K, and the divertor cryopumps at a temperature of 3.7 K. In addition it shall provide refrigeration for the thermal shields at 80 K and deliver helium at 50 K for the current leads. The equivalent refrigeration capacity at 4.5 K will be about 10 kW. The refrigeration process has to be optimised for different operation modes. During the day, in plasma operation state, the refrigerator will cope with the pulsed heat loads which may increase up to 100% of the average power, representing a big challenge compared to other tokamaks. Fast discharge quenches of the magnets, the impact from baking of the vacuum vessel, cool down and warm up modes are presented from the cryogenic system point of view and their impact on the cryogenic design is described.

Effects of MHD instabilities on neutral beam current drive

NASA Astrophysics Data System (ADS)

Podestà, M.; Gorelenkova, M.; Darrow, D. S.; Fredrickson, E. D.; Gerhardt, S. P.; White, R. B.

2015-05-01

Neutral beam injection (NBI) is one of the primary tools foreseen for heating, current drive (CD) and q-profile control in future fusion reactors such as ITER and a Fusion Nuclear Science Facility. However, fast ions from NBI may also provide the drive for energetic particle-driven instabilities (e.g. Alfvénic modes (AEs)), which in turn redistribute fast ions in both space and energy, thus hampering the control capabilities and overall efficiency of NB-driven current. Based on experiments on the NSTX tokamak (M. Ono et al 2000 Nucl. Fusion 40 557), the effects of AEs and other low-frequency magneto-hydrodynamic instabilities on NB-CD efficiency are investigated. A new fast ion transport model, which accounts for particle transport in phase space as required for resonant AE perturbations, is utilized to obtain consistent simulations of NB-CD through the tokamak transport code TRANSP. It is found that instabilities do indeed reduce the NB-driven current density over most of the plasma radius by up to ∼50%. Moreover, the details of the current profile evolution are sensitive to the specific model used to mimic the interaction between NB ions and instabilities. Implications for fast ion transport modeling in integrated tokamak simulations are briefly discussed.

Pushing Particles with Waves: Current Drive and α-Channeling

DOE PAGES

FISCH, Nathaniel J.

2016-01-01

It can be advantageous to push particles with waves in tokamaks or other magnetic confinement devices, relying on wave-particle resonances to accomplish specific goals. Waves that damp on electrons or ions in toroidal fusion devises can drive currents if the waves are launched with toroidal asymmetry. Theses currents are important for tokamaks, since they operate in the absence of an electric field with curl, enabling steady state operation. The lower hybrid wave and the electron cyclotron wave have been demonstrated to drive significant currents. Non-inductive current also stabilizes deleterious tearing modes. Waves can also be used to broker the energymore » transfer between energetic alpha particles and the background plasma. Alpha particles born through fusion reactions in a tokamak reactor tend to slow down on electrons, but that could take up to hundreds of milliseconds. Before that happens, the energy in these alpha particles can destabilize on collisionless timescales toroidal Alfven modes and other waves, in a way deleterious to energy confinement. However, it has been speculated that this energy might be instead be channeled instead into useful energy, that heats fuel ions or drives current. Furthermore, an important question is the extent to which these effects can be accomplished together.« less

Effects of MHD instabilities on neutral beam current drive

DOE PAGES

Podestà, M.; Gorelenkova, M.; Darrow, D. S.; ...

2015-04-17

One of the primary tools foreseen for heating, current drive (CD) and q-profile control in future fusion reactors such as ITER and a Fusion Nuclear Science Facility is the neutral beam injection (NBI). However, fast ions from NBI may also provide the drive for energetic particle-driven instabilities (e.g. Alfvénic modes (AEs)), which in turn redistribute fast ions in both space and energy, thus hampering the control capabilities and overall efficiency of NB-driven current. Based on experiments on the NSTX tokamak (M. Ono et al 2000 Nucl. Fusion 40 557), the effects of AEs and other low-frequency magneto-hydrodynamic instabilities on NB-CDmore » efficiency are investigated. When looking at the new fast ion transport model, which accounts for particle transport in phase space as required for resonant AE perturbations, is utilized to obtain consistent simulations of NB-CD through the tokamak transport code TRANSP. It is found that instabilities do indeed reduce the NB-driven current density over most of the plasma radius by up to ~50%. Moreover, the details of the current profile evolution are sensitive to the specific model used to mimic the interaction between NB ions and instabilities. Finally, implications for fast ion transport modeling in integrated tokamak simulations are briefly discussed.« less

Fracture toughness testing of Linde 1092 reactor vessel welds in the transition range using Charpy-sized specimens

DOE Office of Scientific and Technical Information (OSTI.GOV)

Pavinich, W.A.; Yoon, K.K.; Hour, K.Y.

1999-10-01

The present reference toughness method for predicting the change in fracture toughness can provide over estimates of these values because of uncertainties in initial RT{sub NDT} and shift correlations. It would be preferable to directly measure fracture toughness. However, until recently, no standard method was available to characterize fracture toughness in the transition range. ASTM E08 has developed a draft standard that shows promise for providing lower bound transition range fracture toughness using the master curve approach. This method has been successfully implemented using 1T compact fracture specimens. Combustion Engineering reactor vessel surveillance programs do not have compact fracture specimens.more » Therefore, the CE Owners Group developed a program to validate the master curve method for Charpy-sized and reconstituted Charpy-sized specimens for future application on irradiated specimens. This method was validated for Linde 1092 welds using unirradiated Charpy-sized and reconstituted Charpy-sized specimens by comparison of results with those from compact fracture specimens.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dudek, L.; Chrzanowski, J.; Heitzenroeder, P.

The National Compact Stellarator Experiment (NCSX) has been under construction at the Princeton Plasma Physics Laboratory (PPPL) in partnership with the Oak Ridge National Laboratory (ORNL). The stellarator core is designed to produce a compact 3D plasma that combines stellarator and tokamak physics advantages. The complex geometry and tight fabrication tolerances of NCSX create some unique engineering and assembly challenges. The NCSX project was cancelled in May 2008; construction activities are presently being phased out in an orderly fashion. This paper will describe the progress of the fabrication and assembly activities of NCSX. Completion of the coil fabrication is onmore » track for the summer of 2008. All three of the vacuum vessel 120 degrees sections have been delivered. Assembly of vacuum vessel services began in May 2006 and is now complete. Assembly of the modular coils into 3-packs for safe storage is presently underway. (C) 2008 Elsevier B.V. All rights reserved.« less

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

NASA Astrophysics Data System (ADS)

Lyublinski, I. E.; Vertkov, A. V.; Semenov, V. V.

2016-12-01

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vessel elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550-600°C. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18-20 MW/m2. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.

Comparative analysis of the possibility of applying low-melting metals with the capillary-porous system in tokamak conditions

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lyublinski, I. E., E-mail: lyublinski@yandex.ru; Vertkov, A. V., E-mail: avertkov@yandex.ru; Semenov, V. V., E-mail: darkfenix2006@mail.ru

2016-12-15

The use of capillary-porous systems (CPSs) with liquid Li, Ga, and Sn is considered as an alternative for solving the problem of creating plasma-facing elements (PFEs) of the fusion neutron source (FNS) and the DEMO-type reactor. The main advantages of CPSs with liquid metal compared with hard materials are their stability with respect to the degradation of properties in tokamak conditions and capability of surface self-restoration. The evaluation of applicability of liquid metals is performed on the basis of the analysis of their physical and chemical properties, the interaction with the tokamak plasma, and constructive and process features of in-vesselmore » elements with CPSs implementing the application of these metals in a tokamak. It is shown that the upper limit of the PFE working temperature for all low-melting metals under consideration lies in the range of 550–600°Ð¡. The decisive factor for PFEs with Li is the limitation on the admissible atomic flux into plasma, while for those with Ga and Sn it is the corrosion resistance of construction materials. The upper limit of thermal loads in the steady-state operating mode for the considered promising PFE design with the use of Li, Ga, and Sn is close to 18–20 MW/m{sup 2}. It is seen from the analysis that the use of metals with a low equilibrium vapor pressure of (Ga, Sn) gives no gain in extension of the region of admissible working temperatures of PFEs. However, with respect to the totality of properties, the possibility of implementing the self-restoration and stabilization effect of the liquid surface, the influence on the plasma discharge parameters, and the ability to protect the PFE surface in conditions of plasma perturbations and disruption, lithium is the most attractive liquid metal to create CPS-based PFEs for the tokamak.« less

High-intensity power-resolved radiation imaging of an operational nuclear reactor.

PubMed

Beaumont, Jonathan S; Mellor, Matthew P; Villa, Mario; Joyce, Malcolm J

2015-10-09

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors.

High-intensity power-resolved radiation imaging of an operational nuclear reactor

PubMed Central

Beaumont, Jonathan S.; Mellor, Matthew P.; Villa, Mario; Joyce, Malcolm J.

2015-01-01

Knowledge of the neutron distribution in a nuclear reactor is necessary to ensure the safe and efficient burnup of reactor fuel. Currently these measurements are performed by in-core systems in what are extremely hostile environments and in most reactor accident scenarios it is likely that these systems would be damaged. Here we present a compact and portable radiation imaging system with the ability to image high-intensity fast-neutron and gamma-ray fields simultaneously. This system has been deployed to image radiation fields emitted during the operation of a TRIGA test reactor allowing a spatial visualization of the internal reactor conditions to be obtained. The imaged flux in each case is found to scale linearly with reactor power indicating that this method may be used for power-resolved reactor monitoring and for the assay of ongoing nuclear criticalities in damaged nuclear reactors. PMID:26450669

A compact spin-exchange optical pumping system for 3He polarization based on a solenoid coil, a VBG laser diode, and a cosine theta RF coil

NASA Astrophysics Data System (ADS)

Lee, Sungman; Kim, Jongyul; Moon, Myung Kook; Lee, Kye Hong; Lee, Seung Wook; Ino, Takashi; Skoy, Vadim R.; Lee, Manwoo; Kim, Guinyun

2013-02-01

For use as a neutron spin polarizer or analyzer in the neutron beam lines of the HANARO (High-flux Advanced Neutron Application ReactOr) nuclear research reactor, a 3He polarizer was designed based on both a compact solenoid coil and a VBG (volume Bragg grating) diode laser with a narrow spectral linewidth of 25 GHz. The nuclear magnetic resonance (NMR) signal was measured and analyzed using both a built-in cosine radio-frequency (RF) coil and a pick-up coil. Using a neutron transmission measurement, we estimated the polarization ratio of the 3He cell as 18% for an optical pumping time of 8 hours.

Neutral Beam Development for the Lockheed Martin Compact Fusion Reactor

NASA Astrophysics Data System (ADS)

Ebersohn, Frans; Sullivan, Regina

2017-10-01

The Compact Fusion Reactor project at Lockheed Martin Skunk Works is developing a neutral beam injection system for plasma heating. The neutral beam plasma source consists of a high current lanthanum hexaboride (LaB6) hollow cathode which drives an azimuthal cusp discharge similar to gridded ion thrusters. The beam is extracted with a set of focusing grids and is then neutralized in a chamber pumped with Titanium gettering. The design, testing, and analyses of individual components are presented along with the most current full system results. The goal of this project is to advance in-house neutral beam expertise at Lockheed Martin to aid in operation, procurement, and development of neutral beam technology. ©2017 Lockheed Martin Corporation. All Rights Reserved.

THERMAL FISSION REACTOR COMPOSITIONS AND METHOD OF FABRICATING SAME

DOEpatents

Blainey, A.

1959-10-01

A body is presented for use in a thermal fission reactor comprising a sintered compressed mass of a substance of the group consisting of uranium, thorium, and oxides and carbides of uranium and thorium, enclosed in an envelope of a sintered, compacted, heat-conductive material of the group consisting of beryllium, zirconium, and oxides and carbides of beryllium and zirconium.

Fusion energy: Status and prospects

NASA Astrophysics Data System (ADS)

Salomaa, Rainer

A review of the present state of the international fusion research is given. In the largest tokamak devices (JET, TFTR, JT-60) fusion relevant temperatures are routinely obtained and the scientific feasibility of plasma confinement has been demonstrated. Plans concerning the next step are described. A critical view is presented on questions as to what extent the generic advantages of fusion (availability, sufficiency, safety, environmental acceptability, etc.) can be exploited in a practical power reactor where the formidable technological problems call for compromises.

Critical threshold behavior for steady-state internal transport barriers in burning plasmas.

PubMed

García, J; Giruzzi, G; Artaud, J F; Basiuk, V; Decker, J; Imbeaux, F; Peysson, Y; Schneider, M

2008-06-27

Burning tokamak plasmas with internal transport barriers are investigated by means of integrated modeling simulations. The barrier sustainment in steady state, differently from the barrier formation process, is found to be characterized by a critical behavior, and the critical number of the phase transition is determined. Beyond a power threshold, alignment of self-generated and noninductively driven currents occurs and steady state becomes possible. This concept is applied to simulate a steady-state scenario within the specifications of the International Thermonuclear Experimental Reactor.

SoLid: Search for Oscillations with Lithium-6 Detector at the SCK-CEN BR2 reactor

NASA Astrophysics Data System (ADS)

Ban, G.; Beaumont, W.; Buhour, J. M.; Coupé, B.; Cucoanes, A. S.; D'Hondt, J.; Durand, D.; Fallot, M.; Fresneau, S.; Giot, L.; Guillon, B.; Guilloux, G.; Janssen, X.; Kalcheva, S.; Koonen, E.; Labare, M.; Moortgat, C.; Pronost, G.; Raes, L.; Ryckbosch, D.; Ryder, N.; Shitov, Y.; Vacheret, A.; Van Mulders, P.; Van Remortel, N.; Weber, A.; Yermia, F.

2016-04-01

Sterile neutrinos have been considered as a possible explanation for the recent reactor and Gallium anomalies arising from reanalysis of reactor flux and calibration data of previous neutrino experiments. A way to test this hypothesis is to look for distortions of the anti-neutrino energy caused by oscillation from active to sterile neutrino at close stand-off (˜ 6- 8m) of a compact reactor core. Due to the low rate of anti-neutrino interactions the main challenge in such measurement is to control the high level of gamma rays and neutron background. The SoLid experiment is a proposal to search for active-to-sterile anti-neutrino oscillation at very short baseline of the SCK•CEN BR2 research reactor. This experiment uses a novel approach to detect anti-neutrino with a highly segmented detector based on Lithium-6. With the combination of high granularity, high neutron-gamma discrimination using 6LiF:ZnS(Ag) and precise localization of the Inverse Beta Decay products, a better experimental sensitivity can be achieved compared to other state-of-the-art technology. This compact system requires minimum passive shielding allowing for very close stand off to the reactor. The experimental set up of the SoLid experiment and the BR2 reactor will be presented. The new principle of neutrino detection and the detector design with expected performance will be described. The expected sensitivity to new oscillations of the SoLid detector as well as the first measurements made with the 8 kg prototype detector deployed at the BR2 reactor in 2013-2014 will be reported.

Liquid metal magnetohydrodynamics

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lielpeteris, J.; Moreau, R.

1989-01-01

Liquid metal MHD is the subject of this book. It is of central importance in fields like metals processing, energy conversion, nuclear engineering (fast breeders or fusion reactors), geomagnetism and astrophysics. In some circumstances fluid flow phenomena are controlled by an existing magnetic field; the melts in induction furnaces or the liquid metal blanket around future tokamak fusion reactors being significant examples. In other cases the application of an external magnetic field (or of an electric current) may generate drastic modifications in the fluid motion and in the transfer rates; such effects may be used to develop new technologies (electromagneticmore » shaping) or to improve existing techniques (electromagnetic stirring in continuous casting). In the core of the Earth, fluid motion and magnetic fields are both present and their interaction governs important phenomena.« less

Postirradiation thermocyclic loading of ferritic-martensitic structural materials

NASA Astrophysics Data System (ADS)

Belyaeva, L.; Orychtchenko, A.; Petersen, C.; Rybin, V.

Thermonuclear fusion reactors of the Tokamak-type will be unique power engineering plants to operate in thermocyclic mode only. Ferritic-martensitic stainless steels are prime candidate structural materials for test blankets of the ITER fusion reactor. Beyond the radiation damage, thermomechanical cyclic loading is considered as the most detrimental lifetime limiting phenomenon for the above structure. With a Russian and a German facility for thermal fatigue testing of neutron irradiated materials a cooperation has been undertaken. Ampule devices to irradiate specimens for postirradiation thermal fatigue tests have been developed by the Russian partner. The irradiation of these ampule devices loaded with specimens of ferritic-martensitic steels, like the European MANET-II, the Russian 05K12N2M and the Japanese Low Activation Material F82H-mod, in a WWR-M-type reactor just started. A description of the irradiation facility, the qualification of the ampule device and the modification of the German thermal fatigue facility will be presented.

A mini-cavity probe reactor.

NASA Technical Reports Server (NTRS)

Hyland, R. E.

1971-01-01

The mini-cavity reactor is a rocket engine concept which combines the high specific impulse from a central gaseous fueled cavity (0.6 m diam) and NERVA type fuel elements in a driver region that is external to a moderator-reflector zone to produce a compact light weight reactor. The overall dimension including a pressure vessel that is located outside of the spherical reactor is approximately 1.21 m in diameter. Specific impulses up to 2000 sec are obtainable for 220 to 890 N of thrust with pressures less than 1000 atm. Powerplant weights including a radiator for disposing of the power in the driver region are between 4600 and 32,000 kg - less than payloads of the shuttle. This reactor could also be used as a test reactor for gas-core, MHD, breeding and materials research.

Analysis of a boron-carbide-drum-controlled critical reactor experiment

NASA Technical Reports Server (NTRS)

Mayo, W. T.

1972-01-01

In order to validate methods and cross sections used in the neutronic design of compact fast-spectrum reactors for generating electric power in space, an analysis of a boron-carbide-drum-controlled critical reactor was made. For this reactor the transport analysis gave generally satisfactory results. The calculated multiplication factor for the most detailed calculation was only 0.7-percent Delta k too high. Calculated reactivity worth of the control drums was $11.61 compared to measurements of $11.58 by the inverse kinetics methods and $11.98 by the inverse counting method. Calculated radial and axial power distributions were in good agreement with experiment.

Non-inductively driven tokamak plasmas at near-unity β t in the Pegasus toroidal experiment

DOE PAGES

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.; ...

2018-03-14

Amore » major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓ i , high elongation κ , and high toroidal and normalized beta ( β t and β N ) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓ i . The low aspect ratio ( R 0 / a ~ 1.2 ) of Pegasus allows access to high κ and high normalized plasma currents I N = I p / a B T > 14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high β t plasmas. Equilibrium analyses indicate that β t up to ~100% is achieved. Finally, these high β t discharges disrupt at the ideal no-wall β limit at β N ~ 7. « less

Edge multi-energy soft x-ray diagnostic in Experimental Advanced Superconducting Tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Li, Y. L.; Xu, G. S.; Wan, B. N.

A multi-energy soft x-ray (ME-SXR) diagnostic has been built for electron temperature profile in the edge plasma region in Experimental Advanced Superconducting Tokamak (EAST) after two rounds of campaigns. Originally, five preamplifiers were mounted inside the EAST vacuum vessel chamber attached to five vertically stacked compact diode arrays. A custom mechanical structure was designed to protect the detectors and electronics under constraints of the tangential field of view for plasma edge and the allocation of space. In the next experiment, the mechanical structure was redesigned with a barrel structure to absolutely isolate it from the vacuum vessel. Multiple shielding structuresmore » were mounted at the pinhole head to protect the metal foils from lithium coating. The pre-amplifiers were moved to the outside of the vacuum chamber to avoid introducing interference. Twisted copper cooling tube was embedded into the back-shell near the diode to limit the temperature of the preamplifiers and diode arrays during vacuum vessel baking when the temperature reached 150 °C. Electron temperature profiles were reconstructed from ME-SXR measurements using neural networks.« less

Non-inductively driven tokamak plasmas at near-unity β t in the Pegasus toroidal experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Reusch, Joshua A.; Bodner, Grant M.; Bongard, Michael W.

Amore » major goal of the spherical tokamak (ST) research program is accessing a state of low internal inductance ℓ i , high elongation κ , and high toroidal and normalized beta ( β t and β N ) without solenoidal current drive. Local helicity injection (LHI) in the Pegasus ST [Garstka et al., Nucl. Fusion 46, S603 (2006)] provides non-solenoidally driven plasmas that exhibit these characteristics. LHI utilizes compact, edge-localized current sources for plasma startup and sustainment. It results in hollow current density profiles with low ℓ i . The low aspect ratio ( R 0 / a ~ 1.2 ) of Pegasus allows access to high κ and high normalized plasma currents I N = I p / a B T > 14 ). Magnetic reconnection during LHI provides auxiliary ion heating. Together, these features provide access to very high β t plasmas. Equilibrium analyses indicate that β t up to ~100% is achieved. Finally, these high β t discharges disrupt at the ideal no-wall β limit at β N ~ 7. « less

Intermittent bursts induced by double tearing mode reconnection

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wei, Lai; Wang, Zheng-Xiong, E-mail: zxwang@dlut.edu.cn

Reversed magnetic shear (RMS) configuration is assumed to be the steady-state operation scenario for the future advanced tokamaks like International Thermonuclear Experimental Reactor. In this work, we numerically discover a phenomenon of violent intermittent bursts induced by self-organized double tearing mode (DTM) reconnection in the RMS configuration during the very long evolution, which may continuously lead to annular sawtooth crashes and thus badly impact the desired steady-state operation of the future advanced RMS tokamaks. The key process of the intermittent bursts in the off-axis region is similar to that of the typical sawtooth relaxation oscillation in the positive magnetic shearmore » configuration. It is interestingly found that in the decay phase of the DTM reconnection, the zonal field significantly counteracts equilibrium field to make the magnetic shear between the two rational surfaces so weak that the residual self-generated vortices of the previous DTM burst are able to trigger a reverse DTM reconnection by curling the field lines.« less

Dynamic simulations for preparing the acceptance test of JT-60SA cryogenic system

NASA Astrophysics Data System (ADS)

Cirillo, R.; Hoa, C.; Michel, F.; Poncet, J. M.; Rousset, B.

2016-12-01

Power generation in the future could be provided by thermo-nuclear fusion reactors like tokamaks. There inside, the fusion reaction takes place thanks to the generation of plasmas at hundreds of millions of degrees that must be confined magnetically with superconductive coils, cooled down to around 4.5 K. Within this frame, an experimental tokamak device, JT-60SA is currently under construction in Naka (Japan). The plasma works cyclically and the coil system is subject to pulsed heat loads. In order to size the refrigerator close to the average power and hence optimizing investment and operational costs, measures have to be taken to smooth the heat load. Here we present a dynamic model of the JT-60SA's Auxiliary Cold box (ACB) for preparing the acceptance tests of the refrigeration system planned in 2016 in Naka. The aim of this study is to simulate the pulsed load scenarios using different process controls. All the simulations have been performed with EcosimPro® and the associated cryogenic library: CRYOLIB.

Gyrokinetic theory of turbulent acceleration and momentum conservation in tokamak plasmas

NASA Astrophysics Data System (ADS)

Lu, WANG; Shuitao, PENG; P, H. DIAMOND

2018-07-01

Understanding the generation of intrinsic rotation in tokamak plasmas is crucial for future fusion reactors such as ITER. We proposed a new mechanism named turbulent acceleration for the origin of the intrinsic parallel rotation based on gyrokinetic theory. The turbulent acceleration acts as a local source or sink of parallel rotation, i.e., volume force, which is different from the divergence of residual stress, i.e., surface force. However, the order of magnitude of turbulent acceleration can be comparable to that of the divergence of residual stress for electrostatic ion temperature gradient (ITG) turbulence. A possible theoretical explanation for the experimental observation of electron cyclotron heating induced decrease of co-current rotation was also proposed via comparison between the turbulent acceleration driven by ITG turbulence and that driven by collisionless trapped electron mode turbulence. We also extended this theory to electromagnetic ITG turbulence and investigated the electromagnetic effects on intrinsic parallel rotation drive. Finally, we demonstrated that the presence of turbulent acceleration does not conflict with momentum conservation.

Vertical Position and Current Profile Measurements by Faraday-effect Polarimetry On EAST tokamak

NASA Astrophysics Data System (ADS)

Ding, Weixing; Liu, H. Q.; Jie, Y. X.; Brower, D. L.; Qian, J. P.; Zou, Z. Y.; Lian, H.; Wang, S. X.; Luo, Z. P.; Xiao, B. J.; Ucla Team; Asipp Team

2017-10-01

A primary goal for ITER and prospective fusion power reactors is to achieve controlled long-pulse/steady-state burning plasmas. For elongated divertor plasmas, both the vertical position and current profile have to be precisely controlled to optimize performance and prevent disruptions. An eleven-channel laser-based POlarimeter-INTerferometer (POINT) system has been developed for measuring the internal magnetic field in the EAST tokamak and can be used to obtain the plasma current profile and vertical position. Current profiles are determined from equilibrium reconstruction including internal magnetic field measurements as internal constraints. Horizontally-viewing chords at/near the mid-plane allow us to determine plasma vertical position non-inductively with subcentimeter spatial resolution and time response up to 1 s. The polarimeter-based position measurement, which does not require equilibrium reconstruction, is benchmarked against conventional flux loop measurements and can be exploited for feedback control. Work supported by US DOE through Grants No. DE-FG02-01ER54615 and No. DC-SC0010469.

Current drive with combined electron cyclotron wave and high harmonic fast wave in tokamak plasmas

NASA Astrophysics Data System (ADS)

Li, J. C.; Gong, X. Y.; Dong, J. Q.; Wang, J.; Zhang, N.; Zheng, P. W.; Yin, C. Y.

2016-12-01

The current driven by combined electron cyclotron wave (ECW) and high harmonic fast wave is investigated using the GENRAY/CQL3D package. It is shown that no significant synergetic current is found in a range of cases with a combined ECW and fast wave (FW). This result is consistent with a previous study [Harvey et al., in Proceedings of IAEA TCM on Fast Wave Current Drive in Reactor Scale Tokamaks (Synergy and Complimentarily with LHCD and ECRH), Arles, France, IAEA, Vienna, 1991]. However, a positive synergy effect does appear with the FW in the lower hybrid range of frequencies. This positive synergy effect can be explained using a picture of the electron distribution function induced by the ECW and a very high harmonic fast wave (helicon). The dependence of the synergy effect on the radial position of the power deposition, the wave power, the wave frequency, and the parallel refractive index is also analyzed, both numerically and physically.

Intermittent bursts induced by double tearing mode reconnection

NASA Astrophysics Data System (ADS)

Wei, Lai; Wang, Zheng-Xiong

2014-06-01

Reversed magnetic shear (RMS) configuration is assumed to be the steady-state operation scenario for the future advanced tokamaks like International Thermonuclear Experimental Reactor. In this work, we numerically discover a phenomenon of violent intermittent bursts induced by self-organized double tearing mode (DTM) reconnection in the RMS configuration during the very long evolution, which may continuously lead to annular sawtooth crashes and thus badly impact the desired steady-state operation of the future advanced RMS tokamaks. The key process of the intermittent bursts in the off-axis region is similar to that of the typical sawtooth relaxation oscillation in the positive magnetic shear configuration. It is interestingly found that in the decay phase of the DTM reconnection, the zonal field significantly counteracts equilibrium field to make the magnetic shear between the two rational surfaces so weak that the residual self-generated vortices of the previous DTM burst are able to trigger a reverse DTM reconnection by curling the field lines.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Bentz, L.K.; Bender, D.S.

This report contains estimates for current and projected population distributions for a 50-mile radius area around the Princeton Plasma Physics Laboratory (PPPL) in central New Jersey. It was prepared as supplemental information for an environmental assessment for the proposed Compact Ignition Tokamak facility at PPPL. The report contains estimates for the 1985 population, as well as projections for 1995, 2000, 2005 and 2010. Estimates are provided for municipalities and counties in the area, as well as for a sectorized annular grid centered at PPPL. Sectorized maps are included. An appendix contains technical details on the methodology used. 17 refs., 10more » figs., 13 tabs.« less

Initial Gamma Spectrometry Examination of the AGR-3/4 Irradiation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason M.; Demkowicz, Paul A.; Stempien, John D.

2016-11-01

The initial results from gamma spectrometry examination of the different components from the combined third and fourth US Advanced Gas Reactor Fuel Development TRISO-coated particle fuel irradiation tests (AGR-3/4) have been analyzed. This experiment was designed to provide information about in-pile fission product migration. In each of the 12 capsules, a single stack of four compacts with designed-to-fail particles surrounded by two graphitic diffusion rings (inner and outer) and a graphite sink were irradiated in the Idaho National Laboratory’s Advanced Test Reactor. Gamma spectrometry has been used to evaluate the gamma-emitting fission product inventory of compacts from the irradiation andmore » evaluate the burnup of these compacts based on the activity of the radioactive cesium isotopes (Cs-134 and Cs-137) in the compacts. Burnup from gamma spectrometry compares well with predicted burnup from simulations. Additionally, inner and outer rings were also examined by gamma spectrometry both to evaluate the fission product inventory and the distribution of gamma-emitting fission products within the rings using gamma emission computed tomography. The cesium inventory of the scanned rings compares acceptably well with the expected inventory from fission product transport modeling. The inventory of the graphite fission product sinks is also being evaluated by gamma spectrometry.« less

FALCON reactor-pumped laser description and program overview

DOE Office of Scientific and Technical Information (OSTI.GOV)

None

1989-12-01

The FALCON (Fission Activated Laser CONcept) reactor-pumped laser program at Sandia National Laboratories is examining the feasibility of high-power systems pumped directly by the energy from a nuclear reactor. In this concept we use the highly energetic fission fragments from neutron induced fission to excite a large volume laser medium. This technology has the potential to scale to extremely large optical power outputs in a primarily self-powered device. A laser system of this type could also be relatively compact and capable of long run times without refueling.

Comparison study of toroidal-field divertors for a compact reversed-field pinch reactor

NASA Astrophysics Data System (ADS)

Bathke, C. G.; Krakowski, R. A.; Miller, R. L.

Two divertor configurations for the Compact Reversed-Field Pinch Reactor (CRFPR) based on diverting the minority (toroidal) field have been reported. A critical factor in evaluating the performance of both poloidally symmetric and bundle divertor configurations is the accurate determination of the divertor connection length and the monitoring of magnetic islands introduced by the divertors, the latter being a three-dimensional effect. To this end the poloidal-field, toroidal-field, and divertor coils and the plasma currents are simulated in three dimensions for field-line trackings in both the divertor channel and the plasma-edge regions. The results of this analysis indicate a clear preference for the poloidally symmetric toroidal-field divertor. Design modifications to the limiter-based CRFPR design that accommodate this divertor are presented.

A quasilinear operator retaining magnetic drift effects in tokamak geometry

NASA Astrophysics Data System (ADS)

Catto, Peter J.; Lee, Jungpyo; Ram, Abhay K.

2017-12-01

The interaction of radio frequency waves with charged particles in a magnetized plasma is usually described by the quasilinear operator that was originally formulated by Kennel & Engelmann (Phys. Fluids, vol. 9, 1966, pp. 2377-2388). In their formulation the plasma is assumed to be homogenous and embedded in a uniform magnetic field. In tokamak plasmas the Kennel-Engelmann operator does not capture the magnetic drifts of the particles that are inherent to the non-uniform magnetic field. To overcome this deficiency a combined drift and gyrokinetic derivation is employed to derive the quasilinear operator for radio frequency heating and current drive in a tokamak with magnetic drifts retained. The derivation requires retaining the magnetic moment to higher order in both the unperturbed and perturbed kinetic equations. The formal prescription for determining the perturbed distribution function then follows a novel procedure in which two non-resonant terms must be evaluated explicitly. The systematic analysis leads to a diffusion equation that is compact and completely expressed in terms of the drift kinetic variables. The equation is not transit averaged, and satisfies the entropy principle, while retaining the full poloidal angle variation without resorting to Fourier decomposition. As the diffusion equation is in physical variables, it can be implemented in any computational code. In the Kennel-Engelmann formalism, the wave-particle resonant delta function is either for the Landau resonance or the Doppler shifted cyclotron resonance. In the combined gyro and drift kinetic approach, a term related to the magnetic drift modifies the resonance condition.

Assessment of quasi-linear effect of RF power spectrum for enabling lower hybrid current drive in reactor plasmas

NASA Astrophysics Data System (ADS)

Cesario, Roberto; Cardinali, Alessandro; Castaldo, Carmine; Amicucci, Luca; Ceccuzzi, Silvio; Galli, Alessandro; Napoli, Francesco; Panaccione, Luigi; Santini, Franco; Schettini, Giuseppe; Tuccillo, Angelo Antonio

2017-10-01

The main research on the energy from thermonuclear fusion uses deuterium plasmas magnetically trapped in toroidal devices. To suppress the turbulent eddies that impair thermal insulation and pressure tight of the plasma, current drive (CD) is necessary, but tools envisaged so far are unable accomplishing this task while efficiently and flexibly matching the natural current profiles self-generated at large radii of the plasma column [1-5]. The lower hybrid current drive (LHCD) [6] can satisfy this important need of a reactor [1], but the LHCD system has been unexpectedly mothballed on JET. The problematic extrapolation of the LHCD tool at reactor graded high values of, respectively, density and temperatures of plasma has been now solved. The high density problem is solved by the FTU (Frascati Tokamak Upgrade) method [7], and solution of the high temperature one is presented here. Model results based on quasi-linear (QL) theory evidence the capability, w.r.t linear theory, of suitable operating parameters of reducing the wave damping in hot reactor plasmas. Namely, using higher RF power densities [8], or a narrower antenna power spectrum in refractive index [9,10], the obstacle for LHCD represented by too high temperature of reactor plasmas should be overcome. The former method cannot be used for routinely, safe antenna operations, Thus, only the latter key is really exploitable in a reactor. The proposed solutions are ultimately necessary for viability of an economic reactor.

NE-213-scintillator-based neutron detection system for diagnostic measurements of energy spectra for neutrons having energies greater than or equal to 0. 8 MeV created during plasma operations at the Princeton Tokamak Fusion Test Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Dickens, J.K.; Hill, N.W.; Hou, F.S.

1985-08-01

A system for making diagnostic measurements of the energy spectra of greater than or equal to 0.8-MeV neutrons produced during plasma operations of the Princeton Tokamak Fusion Test Reactor (TFTR) has been fabricated and tested and is presently in operation in the TFTR Test Cell Basement. The system consists of two separate detectors, each made up of cells containing liquid NE-213 scintillator attached permanently to RCA-8850 photomultiplier tubes. Pulses obtained from each photomultiplier system are amplified and electronically analyzed to identify and separate those pulses due to neutron-induced events in the detector from those due to photon-induced events in themore » detector. Signals from each detector are routed to two separate Analog-to-Digital Converters, and the resulting digitized information, representing: (1) the raw neutron-spectrum data; and (2) the raw photon-spectrum data, are transmited to the CICADA data-acquisition computer system of the TFTR. Software programs have been installed on the CICADA system to analyze the raw data to provide moderate-resolution recreations of the energy spectrum of the neutron and photon fluences incident on the detector during the operation of the TFTR. A complete description of, as well as the operation of, the hardware and software is given in this report.« less

Fabrication methods and anisotropic properties of graphite matrix compacts for use in HTGR

NASA Astrophysics Data System (ADS)

Yeo, Sunghwan; Yun, Jihae; Kim, Sungok; Cho, Moon Sung; Lee, Young-Woo

2018-02-01

This study investigated the anisotropic microstructural, mechanical, and thermal properties of fabricated graphite matrix prismatic compacts for High Temperature Gas Cooled Reactor (HTGR) fuel. When the observed alignment of graphite grains and the coke derived from phenolic resin is in the transverse direction, the result is severely anisotropic thermal properties. Compacts with such orientation in the transverse direction exhibited increases of thermal expansion and conductivity up to 5.8 times and 4.82 times, respectively, more than those in the axial direction. The formation of pores due to the pyrolysis of phenolic resin was observed predominantly on upper region of the fabricated compacts. This anisotropic pore formation created anisotropic Vickers hardness on the planes with different directions.

Optimal Control Techniques for ResistiveWall Modes in Tokamaks

NASA Astrophysics Data System (ADS)

Clement, Mitchell Dobbs Pearson

Tokamaks can excite kink modes that can lock or nearly lock to the vacuum vessel wall, and whose rotation frequencies and growth rates vary in time but are generally inversely proportional to the magnetic flux diffusion time of the vacuum vessel wall. This magnetohydrodynamic (MHD) instability is pressure limiting in tokamaks and is called the Resistive Wall Mode (RWM). Future tokamaks that are expected to operate as fusion reactors will be required to maximize plasma pressure in order to maximize fusion performance. The DIII-D tokamak is equipped with electromagnetic control coils, both inside and outside of its vacuum vessel, which create magnetic fields that are small by comparison to the machine's equilibrium field but are able to dynamically counteract the RWM. Presently for RWM feedback, DIII-D uses its interior control coils using a classical proportional gain only controller to achieve high plasma pressure. Future advanced tokamak designs will not likely have the luxury of interior control coils and a proportional gain algorithm is not expected to be effective with external control coils. The computer code VALEN was designed to calculate the performance of an MHD feedback control system in an arbitrary geometry. VALEN models the perturbed magnetic field from a single MHD instability and its interaction with surrounding conducting structures using a finite element approach. A linear quadratic gaussian (LQG) control, or H 2 optimal control, algorithm based on the VALEN model for RWM feedback was developed for use with DIII-D's external control coil set. The algorithm is implemented on a platform that combines a graphics processing unit (GPU) for real-time control computation with low latency digital input/output control hardware and operates in parallel with the DIII-D Plasma Control System (PCS). Simulations and experiments showed that modern control techniques performed better, using 77% less current, than classical techniques when using coils external to the vacuum vessel for RWM feedback. RWM feedback based on VALEN outperformed a classical control algorithm using external coils to suppress the normalized plasma response to a rotating n=1 perturbation applied by internal coils over a range of frequencies. This study describes the design, development and testing of the GPU based control hardware and algorithm along with its performance during experiment and simulation.

Waste management for different fusion reactor designs

NASA Astrophysics Data System (ADS)

Rocco, Paolo; Zucchetti, Massimo

2000-12-01

Safety and Environmental Assessment of Fusion Power (SEAFP) waste management studies performed up to 1998 concerned three power tokamak designs. In-vessel structural materials consist of V-alloys or low activation martensitic (LAM) steel; tritium-producing materials are Li 2O, Pb-17Li, Li 4SiO 4 with a Be-multiplier; coolants are helium or water. The strategy chosen reduces permanent radwaste by recycling the in-vessel materials and by clearance of the other structures. Limits of the contact dose rate and specific activity of the waste allowing such options are defined accordingly. SEAFP activities for 1999 enlarge the analysis to three additional reactors with in-vessel structures made with SiC/SiC composites. These materials cannot be recycled due to their form and, according to national regulations of E.C. countries, long-lived activation products hinder near-surface burial (NSB).

Tritium environmental transport studies at TFTR

NASA Astrophysics Data System (ADS)

Ritter, P. D.; Dolan, T. J.; Longhurst, G. R.

1993-06-01

Environmental tritium concentrations will be measured near the Tokamak Fusion Test Reactor (TFTR) to help validate dynamic models of tritium transport in the environment. For model validation the database must contain sequential measurements of tritium concentrations in key environmental compartments. Since complete containment of tritium is an operational goal, the supplementary monitoring program should be able to glean useful data from an unscheduled acute release. Portable air samplers will be used to take samples automatically every 4 hours for a week after an acute release, thus obtaining the time resolution needed for code validation. Samples of soil, vegetation, and foodstuffs will be gathered daily at the same locations as the active air monitors. The database may help validate the plant/soil/air part of tritium transport models and enhance environmental tritium transport understanding for the International Thermonuclear Experimental Reactor (ITER).

Japan Report No. 173.

DTIC Science & Technology

1983-06-10

nuclear ship Mutsu . We will in parallel pursue the development of an advanced ma- rine reactor of more compact and ef- ficient design. In the field...velopment include such subjects as the High Temperature Gas Reactor - envisaged for uses other than power generation - and nuclear ship pro- pulsion. We...950’centigrade. In the field of nuclear ship pro- pulsion, we will proceed on our plans for experimental voyages to be undertaken by our first

Chapter 8: Plasma operation and control

NASA Astrophysics Data System (ADS)

ITER Physics Expert Group on Disruptions, Control, Plasma, and MHD; ITER Physics Expert Group on Energetic Particles, Heating, Current and Drive; ITER Physics Expert Group on Diagnostics; ITER Physics Basis Editors

1999-12-01

Wall conditioning of fusion devices involves removal of desorbable hydrogen isotopes and impurities from interior device surfaces to permit reliable plasma operation. Techniques used in present devices include baking, metal film gettering, deposition of thin films of low-Z material, pulse discharge cleaning, glow discharge cleaning, radio frequency discharge cleaning, and in situ limiter and divertor pumping. Although wall conditioning techniques have become increasingly sophisticated, a reactor scale facility will involve significant new challenges, including the development of techniques applicable in the presence of a magnetic field and of methods for efficient removal of tritium incorporated into co-deposited layers on plasma facing components and their support structures. The current status of various approaches is reviewed, and the implications for reactor scale devices are summarized. Creation and magnetic control of shaped and vertically unstable elongated plasmas have been mastered in many present tokamaks. The physics of equilibrium control for reactor scale plasmas will rely on the same principles, but will face additional challenges, exemplified by the ITER/FDR design. The absolute positioning of outermost flux surface and divertor strike points will have to be precise and reliable in view of the high heat fluxes at the separatrix. Long pulses will require minimal control actions, to reduce accumulation of AC losses in superconducting PF and TF coils. To this end, more complex feedback controllers are envisaged, and the experimental validation of the plasma equilibrium response models on which such controllers are designed is encouraging. Present simulation codes provide an adequate platform on which equilibrium response techniques can be validated. Burning plasmas require kinetic control in addition to traditional magnetic shape and position control. Kinetic control refers to measures controlling density, rotation and temperature in the plasma core as well as in plasma periphery and divertor. The planned diagnostics (Chapter 7) serve as sensors for kinetic control, while gas and pellet fuelling, auxiliary power and angular momentum input, impurity injection, and non-inductive current drive constitute the control actuators. For example, in an ignited plasma, core density controls fusion power output. Kinetic control algorithms vary according to the plasma state, e.g. H- or L-mode. Generally, present facilities have demonstrated the kinetic control methods required for a reactor scale device. Plasma initiation - breakdown, burnthrough and initial current ramp - in reactor scale tokamaks will not involve physics differing from that found in present day devices. For ITER, the induced electric field in the chamber will be ~0.3V· m-1 - comparable to that required by breakdown theory but somewhat smaller than in present devices. Thus, a start-up 3MW electron cyclotron heating system will be employed to assure burnthrough. Simulations show that plasma current ramp up and termination in a reactor scale device can follow procedures developed to avoid disruption in present devices. In particular, simulations remain in the stable area of the li-q plane. For design purposes, the resistive V·s consumed during initiation is found, by experiments, to follow the Ejima expression, 0.45μ0 RIp. Advanced tokamak control has two distinct goals. First, control of density, auxiliary power, and inductive current ramping to attain reverse shear q profiles and internal transport barriers, which persist until dissipated by magnetic flux diffusion. Such internal transport barriers can lead to transient ignition. Second, combined use poloidal field shape control with non-inductive current drive and NBI angular momentum injection to create and control steady state, high bootstrap fraction, reverse shear discharges. Active n = 1 magnetic feedback and/or driven rotation will be required to suppress resistive wall modes for steady state plasmas that must operate in the wall stabilized regime for reactor levels of β >= 0.03.

CompactPCI/Linux Platform in FTU Slow Control System

NASA Astrophysics Data System (ADS)

Iannone, F.; Wang, L.; Centioli, C.; Panella, M.; Mazza, G.; Vitale, V.

2004-12-01

In large fusion experiments, such as tokamak devices, there is a common trend for slow control systems. Because of complexity of the plants, the so-called `Standard Model' (SM) in slow control has been adopted on several tokamak machines. This model is based on a three-level hierarchical control: 1) High-Level Control (HLC) with a supervisory function; 2) Medium-Level Control (MLC) to interface and concentrate I/O field equipments; 3) Low-Level Control (LLC) with hard real-time I/O function, often managed by PLCs. FTU control system designed with SM concepts has underwent several stages of developments in its fifteen years duration of runs. The latest evolution was inevitable, due to the obsolescence of the MLC CPUs, based on VME-MOTOROLA 68030 with OS9 operating system. A large amount of C code was developed for that platform to route the data flow from LLC, which is constituted by 24 Westinghouse Numalogic PC-700 PLCs with about 8000 field-points, to HLC, based on a commercial Object-Oriented Real-Time database on Alpha/CompaqTru64 platform. Therefore, we have to look for cost-effective solutions and finally a CompactPCI-Intel x86 platform with Linux operating system was chosen. A software porting has been done, taking into account the differences between OS9 and Linux operating system in terms of Inter/Network Processes Communications and I/O multi-ports serial driver. This paper describes the hardware/software architecture of the new MLC system, emphasizing the reliability and the low costs of the open source solutions. Moreover, a huge amount of software packages available in open source environment will assure a less painful maintenance, and will open the way to further improvements of the system itself.

Evaluation of weldments in type 21-6-9 stainless steel for compact ignition tokamak structural applications, phase 1

NASA Astrophysics Data System (ADS)

Alexander, D. J.; Goodwin, G. M.; Bloom, E. E.

1991-06-01

Primary design considerations for the Compact Ignition Tokamak toroidal field-coil cases are yield strength and toughness in the temperature range from 77 to 300 K. Type 21-6-9 stainless steel, also still known by its original Armco Steel Company trade name Nitronic 40, is the proposed alloy for this application. It has high yield strength and usually adequate base metal toughness, but weldments in thick sections have not been adequately characterized in terms of mechanical properties or hot-cracking propensity. In this study, weldability of the alloy in heavy sections and the mechanical properties of the resultant welds were investigated including tensile yield strength and Charpy V-notch toughness at 77 K and room temperature. Weldments were made in four different base metals using seven different filler metals. None of the weldments showed any indication of hot-cracking problems. All base metals, including weldment heat-affected zones, were found to have adequate strength and impact toughness at both test temperatures. Weld metals, on the other hand, except ERNiCr-3 and ENiCrFe-3, had impact toughnesses of less than 67 J at 77 K. Inconel 82 had an average weld metal impact toughness of over 135 J at 77 K, and although its strength at 77 K is less than that of type 21-6-9 base metal, at this point it is considered to be the first-choice filler metal. Phase 2 of this program will concentrate on composition refinement and process/procedure optimization for the generic ERNiCr-3 composition and will generate a design data base for base and weld metal, including tensile, fracture toughness, and crack growth rate data.

Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

DOE PAGES

Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

2015-11-04

We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO 2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize themore » degradation rate constant.« less

Stacked waveguide reactors with gradient embedded scatterers for high-capacity water cleaning

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ahsan, Syed Saad; Gumus, Abdurrahman; Erickson, David

We present a compact water-cleaning reactor with stacked layers of waveguides containing gradient patterns of optical scatterers that enable uniform light distribution and augmented water-cleaning rates. Previous photocatalytic reactors using immersion, external, or distributive lamps suffer from poor light distribution that impedes scalability. Here, we use an external UV-source to direct photons into stacked waveguide reactors where we scatter the photons uniformly over the length of the waveguide to thin films of TiO 2-catalysts. In conclusion, we also show 4.5 times improvement in activity over uniform scatterer designs, demonstrate a degradation of 67% of the organic dye, and characterize themore » degradation rate constant.« less

REACTOR CONTROL DEVICE

DOEpatents

Graham, R.H.

1962-09-01

A wholly mechanical compact control device is designed for automatically rendering the core of a fission reactor subcritical in response to core temperatures in excess of the design operating temperature limit. The control device comprises an expansible bellows interposed between the base of a channel in a reactor core and the inner end of a fuel cylinder therein which is normally resiliently urged inwardly. The bellows contains a working fluid which undergoes a liquid to vapor phase change at a temperature substantially equal to the design temperature limit. Hence, the bellows abruptiy expands at this limiting temperature to force the fuel cylinder outward and render the core subcritical. The control device is particularly applicable to aircraft propulsion reactor service. (AEC)

3D toroidal physics: testing the boundaries of symmetry breaking

NASA Astrophysics Data System (ADS)

Spong, Don

2014-10-01

Toroidal symmetry is an important concept for plasma confinement; it allows the existence of nested flux surface MHD equilibria and conserved invariants for particle motion. However, perfect symmetry is unachievable in realistic toroidal plasma devices. For example, tokamaks have toroidal ripple due to discrete field coils, optimized stellarators do not achieve exact quasi-symmetry, the plasma itself continually seeks lower energy states through helical 3D deformations, and reactors will likely have non-uniform distributions of ferritic steel near the plasma. Also, some level of designed-in 3D magnetic field structure is now anticipated for most concepts in order to lead to a stable, steady-state fusion reactor. Such planned 3D field structures can take many forms, ranging from tokamaks with weak 3D ELM-suppression fields to stellarators with more dominant 3D field structures. There is considerable interest in the development of unified physics models for the full range of 3D effects. Ultimately, the questions of how much symmetry breaking can be tolerated and how to optimize its design must be addressed for all fusion concepts. Fortunately, significant progress is underway in theory, computation and plasma diagnostics on many issues such as magnetic surface quality, plasma screening vs. amplification of 3D perturbations, 3D transport, influence on edge pedestal structures, MHD stability effects, modification of fast ion-driven instabilities, prediction of energetic particle heat loads on plasma-facing materials, effects of 3D fields on turbulence, and magnetic coil design. A closely coupled program of simulation, experimental validation, and design optimization is required to determine what forms and amplitudes of 3D shaping and symmetry breaking will be compatible with future fusion reactors. The development of models to address 3D physics and progress in these areas will be described. This work is supported both by the US Department of Energy under Contract DE-AC05-00OR22725 with UT-Battelle, LLC and under the US DOE SciDAC GSEP Center.

Approach to ignition of tokamak reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sigmar, D.J.

1981-02-01

Recent transport modeling results for JET, INTOR, and ETF are reviewed and analyzed with respect to existing uncertainties in the underlying physics, the self-consistency of the very large numerical codes, and the margin for ignition. The codes show ignition to occur in ETF/INTOR-sized machines if empirical scaling can be extrapolated to ion temperatures (and beta values) much higher than those presently achieved, if there is no significant impurity accumulation over the first 7 s, and if the known ideal and resistive MHD instabilities remain controllable for the evolving plasma profiles during ignition startup.

Advanced divertor configurations with large flux expansion

NASA Astrophysics Data System (ADS)

Soukhanovskii, V. A.; Bell, R. E.; Diallo, A.; Gerhardt, S.; Kaye, S.; Kolemen, E.; LeBlanc, B. P.; McLean, A.; Menard, J. E.; Paul, S. F.; Podesta, M.; Raman, R.; Ryutov, D. D.; Scotti, F.; Kaita, R.; Maingi, R.; Mueller, D. M.; Roquemore, A. L.; Reimerdes, H.; Canal, G. P.; Labit, B.; Vijvers, W.; Coda, S.; Duval, B. P.; Morgan, T.; Zielinski, J.; De Temmerman, G.; Tal, B.

2013-07-01

Experimental studies of the novel snowflake divertor concept (D. Ryutov, Phys. Plasmas 14 (2007) 064502) performed in the NSTX and TCV tokamaks are reviewed in this paper. The snowflake divertor enables power sharing between divertor strike points, as well as the divertor plasma-wetted area, effective connection length and divertor volumetric power loss to increase beyond those in the standard divertor, potentially reducing heat flux and plasma temperature at the target. It also enables higher magnetic shear inside the separatrix, potentially affecting pedestal MHD stability. Experimental results from NSTX and TCV confirm the predicted properties of the snowflake divertor. In the NSTX, a large spherical tokamak with a compact divertor and lithium-coated graphite plasma-facing components (PFCs), the snowflake divertor operation led to reduced core and pedestal impurity concentration, as well as re-appearance of Type I ELMs that were suppressed in standard divertor H-mode discharges. In the divertor, an otherwise inaccessible partial detachment of the outer strike point with an up to 50% increase in divertor radiation and a peak divertor heat flux reduction from 3-7 MW/m2 to 0.5-1 MW/m2 was achieved. Impulsive heat fluxes due to Type-I ELMs were significantly dissipated in the high magnetic flux expansion region. In the TCV, a medium-size tokamak with graphite PFCs, several advantageous snowflake divertor features (cf. the standard divertor) have been demonstrated: an unchanged L-H power threshold, enhanced stability of the peeling-ballooning modes in the pedestal region (and generally an extended second stability region), as well as an H-mode pedestal regime with reduced (×2-3) Type I ELM frequency and slightly increased (20-30%) normalized ELM energy, resulting in a favorable average energy loss comparison to the standard divertor. In the divertor, ELM power partitioning between snowflake divertor strike points was demonstrated. The NSTX and TCV experiments are providing support for the snowflake divertor as a viable solution for the outstanding tokamak plasma-material interface issues.

Flow tests of a single fuel element coolant channel for a compact fast reactor for space power

NASA Technical Reports Server (NTRS)

Springborn, R. H.

1971-01-01

Water flow tests were conducted on a single-fuel-element cooling channel for a nuclear concept to be used for space power. The tests established a method for measuring coolant flow rate which is applicable to water flow testing of a complete mockup of the reference reactor. The inlet plenum-to-outlet plenum pressure drop, which approximates the overall core pressure drop, was measured and correlated with flow rate. This information can be used for reactor coolant flow and heat transfer calculations. An analytical study of the flow characteristics was also conducted.

Satellite nuclear power station: An engineering analysis

NASA Technical Reports Server (NTRS)

Williams, J. R.; Clement, J. D.; Rosa, R. J.; Kirby, K. D.; Yang, Y. Y.

1973-01-01

A nuclear-MHD power plant system which uses a compact non-breeder reactor to produce power in the multimegawatt range is analyzed. It is shown that, operated in synchronous orbit, the plant would transmit power safely to the ground by a microwave beam. Fuel reprocessing would take place in space, and no radioactive material would be returned to earth. Even the effect of a disastrous accident would have negligible effect on earth. A hydrogen moderated gas core reactor, or a colloid-core, or NERVA type reactor could also be used. The system is shown to approach closely the ideal of economical power without pollution.

Diffusion of cesium and iodine in compressed IG-110 graphite compacts

NASA Astrophysics Data System (ADS)

Carter, L. M.; Brockman, J. D.; Robertson, J. D.; Loyalka, S. K.

2016-08-01

Nuclear graphite grade IG-110 is currently used in the High Temperature Engineering Test Reactor (HTTR) in Japan for certain permanent and replaceable core components, and is a material of interest in general. Therefore, transport parameters for fission products in this material are needed. Measurement of diffusion through pressed compacts of IG-110 graphite is experimentally attractive because they are easy to prepare with homogeneous distributions of fission product surrogates. In this work, we measured diffusion coefficients for Cs and I in pressed compacts made from IG-110 powder in the 1079-1290 K temperature range, and compared them to those obtained in as-received IG-110.

Prediction and realization of ITER-like pedestal pressure in the high- B tokamak Alcator C-Mod

NASA Astrophysics Data System (ADS)

Hughes, Jerry

2017-10-01

Fusion power in a burning plasma will scale as the square of the plasma pressure, which is increased in a straightforward way by increasing magnetic field: Pfus p2 B4 . Experiments on Alcator C-Mod, a compact high- B tokamak, have tested predictive capability for pedestal pressure, at toroidal field BT up to 8T , and poloidal field BP up to 1T . These reactor-like fields enable C-Mod to approach an ITER predicted value of 90kPa . This is expected if, as in the EPED model, the pedestal is constrained by onset of kinetic ballooning modes (KBMs) and peeling-ballooning modes (PMB), yielding a pressure pedestal approximately as pped BT ×BP . One successful path to high confinement on C-Mod is the high-density (ne > 3 ×1020m-3) approach, pursued using enhanced D-alpha (EDAs) H-mode. In EDA H-mode, transport regulates both the pedestal profiles and the core impurity content, holding the pedestal stationary, at just below the PBM stability boundary. We have extended this stationary ELM-suppressed regime to the highest magnetic fields achievable on C-Mod, and used it to approach the maximum pedestal predicted by EPED at high density: pped 60kPa . Another approach to high pressure utilizes a pedestal limited by PBMs at low collisionality, where pressure increases with density and EPED predicts access to a higher ``Super H'' solution for pped. Experiments at reduced density (ne < 2 ×1020m-3) and strong plasma shaping (δ > 0.5) accessed these regimes on C-Mod, producing pedestals with world record pped 80kPa , at Tped 2keV . In both the high and low density approaches, the impact of the pedestal on core performance is substantial. Our exploration of high pedestal regimes yielded a volume-averaged pressure 〈 p 〉 > 2atm , a world record value for a magnetic fusion device. The results hold promise for the projection of pedestal pressure and overall performance of high field burning plasma devices. Supported by U.S. Department of Energy awards DE-FC02-99ER54512, DE-FG02-95ER54309, DE-FC02-06ER54873, DE-AC02-09CH11466, DE-SC0007880 using Alcator C-Mod, a DOE Office of Science User Facility.

Validation of the Physics Analysis used to Characterize the AGR-1 TRISO Fuel Irradiation Test

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sterbentz, James W.; Harp, Jason M.; Demkowicz, Paul A.

2015-05-01

The results of a detailed physics depletion calculation used to characterize the AGR-1 TRISO-coated particle fuel test irradiated in the Advanced Test Reactor (ATR) at the Idaho National Laboratory are compared to measured data for the purpose of validation. The particle fuel was irradiated for 13 ATR power cycles over three calendar years. The physics analysis predicts compact burnups ranging from 11.30-19.56% FIMA and cumulative neutron fast fluence from 2.21?4.39E+25 n/m 2 under simulated high-temperature gas-cooled reactor conditions in the ATR. The physics depletion calculation can provide a full characterization of all 72 irradiated TRISO-coated particle compacts during and post-irradiation,more » so validation of this physics calculation was a top priority. The validation of the physics analysis was done through comparisons with available measured experimental data which included: 1) high-resolution gamma scans for compact activity and burnup, 2) mass spectrometry for compact burnup, 3) flux wires for cumulative fast fluence, and 4) mass spectrometry for individual actinide and fission product concentrations. The measured data are generally in very good agreement with the calculated results, and therefore provide an adequate validation of the physics analysis and the results used to characterize the irradiated AGR-1 TRISO fuel.« less

PBF Reactor Building (PER620). Camera faces south along west wall. ...

Library of Congress Historic Buildings Survey, Historic Engineering Record, Historic Landscapes Survey

PBF Reactor Building (PER-620). Camera faces south along west wall. Gap between native lava rock and concrete basement walls is being backfilled and compacted. Wire mesh protects workers from falling rock. Note penetrations for piping that will carry secondary coolant water to Cooling Tower. Photographer: Holmes. Date: June 15, 1967. INEEL negative no. 67-3665 - Idaho National Engineering Laboratory, SPERT-I & Power Burst Facility Area, Scoville, Butte County, ID

A Compact Nuclear Fusion Reactor for Space Flights

NASA Astrophysics Data System (ADS)

Nastoyashchiy, Anatoly F.

2006-05-01

A small-scale nuclear fusion reactor is suggested based on the concepts of plasma confinement (with a high pressure gas) which have been patented by the author. The reactor considered can be used as a power setup in space flights. Among the advantages of this reactor is the use of a D3He fuel mixture which at burning gives main reactor products — charged particles. The energy balance considerably improves, as synchrotron radiation turn out "captured" in the plasma volume, and dangerous, in the case of classical magnetic confinement, instabilities in the direct current magnetic field configuration proposed do not exist. As a result, the reactor sizes are quite suitable (of the order of several meters). A possibility of making reactive thrust due to employment of ejection of multiply charged ions formed at injection of pellets from some adequate substance into the hot plasma center is considered.

Compact reactor for onboard hydrogen generation

NASA Technical Reports Server (NTRS)

Brabbs, T. A.

1980-01-01

Hydrogen, chemically stored as methanol, is promising internal-combustion fuel. Methanol is readily obtainable from natural products such as wood, compost, or various organic wastes. Steam reformation of methanol as source for hydrogen is relatively simple operation.

Current drive for stability of thermonuclear plasma reactor

NASA Astrophysics Data System (ADS)

Amicucci, L.; Cardinali, A.; Castaldo, C.; Cesario, R.; Galli, A.; Panaccione, L.; Paoletti, F.; Schettini, G.; Spigler, R.; Tuccillo, A.

2016-01-01

To produce in a thermonuclear fusion reactor based on the tokamak concept a sufficiently high fusion gain together stability necessary for operations represent a major challenge, which depends on the capability of driving non-inductive current in the hydrogen plasma. This request should be satisfied by radio-frequency (RF) power suitable for producing the lower hybrid current drive (LHCD) effect, recently demonstrated successfully occurring also at reactor-graded high plasma densities. An LHCD-based tool should be in principle capable of tailoring the plasma current density in the outer radial half of plasma column, where other methods are much less effective, in order to ensure operations in the presence of unpredictably changes of the plasma pressure profiles. In the presence of too high electron temperatures even at the periphery of the plasma column, as envisaged in DEMO reactor, the penetration of the coupled RF power into the plasma core was believed for long time problematic and, only recently, numerical modelling results based on standard plasma wave theory, have shown that this problem should be solved by using suitable parameter of the antenna power spectrum. We show here further information on the new understanding of the RF power deposition profile dependence on antenna parameters, which supports the conclusion that current can be actively driven over a broad layer of the outer radial half of plasma column, thus enabling current profile control necessary for the stability of a reactor.

Physics of the current injection process during localized helicity injection

NASA Astrophysics Data System (ADS)

Hinson, Edward Thomas

An impedance model has been developed for the arc-plasma cathode electron current source used in localized helicity injection tokamak startup. According to this model, a potential double layer (DL) is established between the high-density arc plasma (narc ˜ 1021 m-3) in the electron source, and the less-dense external tokamak edge plasma (nedge ˜ 10 18 m-3) into which current is injected. The DL launches an electron beam at the applied voltage with cross-sectional area close to that of the source aperture: Ainj ≈ 2 cm 2. The injected current, Iinj, increases with applied voltage, Vinj, according to the standard DL scaling, Iinj ˜ V(3/2/ inj), until the more restrictive of two limits to beam density nb arises, producing Iinj ˜ V(1/2/inj), a scaling with beam drift velocity. For low external tokamak edge density nedge, space-charge neutralization of the intense electron beam restricts the injected beam density to nb ˜ nedge. At high Jinj and sufficient edge density, the injected current is limited by expansion of the DL sheath, which leads to nb ˜ narc. Measurements of narc, Iinj , nedge, Vinj, support these predicted scalings, and suggest narc as a viable control actuator for the source impedance. Magnetic probe signals ≈ 300 degrees toroidally from the injection location are consistent with expectations for a gyrating, coherent electron beam with a compact areal cross-section. Technological development of the source has allowed an extension of the favorable Iinj ˜ V(1/2/inj) to higher power without electrical breakdown.

Tritium pellet injector for the tokamak fusion test reactor

NASA Astrophysics Data System (ADS)

Gouge, M. J.; Baylor, L. R.; Combs, S. K.; Fisher, P. W.; Foust, C. R.; Milora, S. L.

The tritium pellet injector (TPI) for the Tokamak Fusion Test Reactor (TFTR) will provide a tritium pellet fueling capability with pellet speeds in the 1- to 3-km/s range for the TFTR deuterium-tritium (D-T) plasma phase. An existing deuterium pellet injector (DPI) was modified at Oak Ridge National Laboratory (ORNL) to provide a four-shot, tritium-compatible, pipe-gun configuration with three upgraded single-stage pneumatic guns and a two-stage light gas gun driver. The TPI was designed for frozen pellets ranging in size from 3 to 4 mm in diameter in arbitrarily programmable firing sequences at tritium pellet speeds up to approximately 1.5 km/s for the three single-stage drivers and 2.5 to 3 km/s for the two-stage driver. Injector operation is controlled by a programmable logic controller (PLC). The new pipe-gun injector assembly was installed in the modified DPI guard vacuum box, and modifications were also made to the internals of the DPI vacuum injection line, including a new pellet diagnostics package. Assembly of these modified parts with existing DPI components was then completed and the TPI was tested at ORNL with deuterium pellets. Results of the testing program at ORNL are described. The TPI has been installed and operated on TFTR in support of the FY-92 deuterium plasma run period. In 1993, the tritium pellet injector will be retrofitted with a D-T fuel manifold and tritium gloveboxes and integrated into TFTR tritium processing systems to provide full tritium pellet capability.

Remote experimental site concept development

NASA Astrophysics Data System (ADS)

Casper, Thomas A.; Meyer, William; Butner, David

1995-01-01

Scientific research is now often conducted on large and expensive experiments that utilize collaborative efforts on a national or international scale to explore physics and engineering issues. This is particularly true for the current US magnetic fusion energy program where collaboration on existing facilities has increased in importance and will form the basis for future efforts. As fusion energy research approaches reactor conditions, the trend is towards fewer large and expensive experimental facilities, leaving many major institutions without local experiments. Since the expertise of various groups is a valuable resource, it is important to integrate these teams into an overall scientific program. To sustain continued involvement in experiments, scientists are now often required to travel frequently, or to move their families, to the new large facilities. This problem is common to many other different fields of scientific research. The next-generation tokamaks, such as the Tokamak Physics Experiment (TPX) or the International Thermonuclear Experimental Reactor (ITER), will operate in steady-state or long pulse mode and produce fluxes of fusion reaction products sufficient to activate the surrounding structures. As a direct consequence, remote operation requiring robotics and video monitoring will become necessary, with only brief and limited access to the vessel area allowed. Even the on-site control room, data acquisition facilities, and work areas will be remotely located from the experiment, isolated by large biological barriers, and connected with fiber-optics. Current planning for the ITER experiment includes a network of control room facilities to be located in the countries of the four major international partners; USA, Russian Federation, Japan, and the European Community.

High heat flux issues for plasma-facing components in fusion reactors

NASA Astrophysics Data System (ADS)

Watson, Robert D.

1993-02-01

Plasma facing components in tokamak fusion reactors are faced with a number of difficult high heat flux issues. These components include: first wall armor tiles, pumped limiters, diverter plates, rf antennae structure, and diagnostic probes. Peak heat fluxes are 15 - 30 MW/m2 for diverter plates, which will operate for 100 - 1000 seconds in future tokamaks. Disruption heat fluxes can approach 100,000 MW/m2 for 0.1 ms. Diverter plates are water-cooled heat sinks with armor tiles brazed on to the plasma facing side. Heat sink materials include OFHC, GlidcopTM, TZM, Mo-41Re, and niobium alloys. Armor tile materials include: carbon fiber composites, beryllium, silicon carbide, tungsten, and molybdenum. Tile thickness range from 2 - 10 mm, and heat sinks are 1 - 3 mm. A twisted tape insert is used to enhance heat transfer and increase the burnout safety margin from critical heat flux limits to 50 - 60 MW/m2 with water at 10 m/s and 4 MPa. Tests using rastered electron beams have shown thermal fatigue failures from cracks at the brazed interface between tiles and the heat sink after only 1000 cycles at 10 - 15 MW/m2. These fatigue lifetimes need to be increased an order of magnitude to meet future requirements. Other critical issues for plasma facing components include: surface erosion from sputtering and disruption erosion, eddy current forces and runaway electron impact from disruptions, neutron damage, tritium retention and release, remote maintenance of radioactive components, corrosion-erosion, and loss-of-coolant accidents.

MERCHANT MARINE SHIP REACTOR

DOEpatents

Mumm, J.F.; North, D.C. Jr.; Rock, H.R.; Geston, D.K.

1961-05-01

A nuclear reactor is described for use in a merchant marine ship. The reactor is of pressurized light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The foregoing design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass.

Merchant Marine Ship Reactor

DOEpatents

Sankovich, M. F.; Mumm, J. F.; North, Jr, D. C.; Rock, H. R.; Gestson, D. K.

1961-05-01

A nuclear reactor for use in a merchant marine ship is described. The reactor is of pressurized, light water cooled and moderated design in which three passes of the water through the core in successive regions of low, intermediate, and high heat generation and downflow in a fuel region are made. The design makes a compact reactor construction with extended core life. The core has an egg-crate lattice containing the fuel elements that are confined between a lower flow baffle and upper grid plate, with the latter serving also as part of a turn- around manifold from which the entire coolant is distributed into the outer fuel elements for the second pass through the core. The inner fuel elements are cooled in the third pass. (AEC)

An Innovative Hybrid Loop-Pool SFR Design and Safety Analysis Methods: Today and Tomorrow

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hongbin Zhang; Haihua Zhao; Vincent Mousseau

2008-04-01

Investment in commercial sodium cooled fast reactor (SFR) power plants will become possible only if SFRs achieve economic competitiveness as compared to light water reactors and other Generation IV reactors. Toward that end, we have launched efforts to improve the economics and safety of SFRs from the thermal design and safety analyses perspectives at Idaho National Laboratory. From the thermal design perspective, an innovative hybrid loop-pool SFR design has been proposed. This design takes advantage of the inherent safety of a pool design and the compactness of a loop design to further improve economics and safety. From the safety analysesmore » perspective, we have initiated an effort to develop a high fidelity reactor system safety code.« less

Fatigue behavior of type 316 stainless steel following neutron irradiation inducing helium

DOE Office of Scientific and Technical Information (OSTI.GOV)

Grossbeck, M.L.; Liu, K.C.

1980-01-01

Since a tokamak fusion reactor operates in a cyclic mode, thermal stresses will result in fatigue in structural components, especially the first wall and blanket. Type 316 stainless steel in the 20% cold-worked condition has been irradiated in the HFIR in order to introduce helium as well as displacement damage. A miniature hourglass specimen was developed for the reactor irradiations and subsequent fully reversed low cycle fatigue testing. For material irradiated and tested at 430/sup 0/C in vacuum to a damage level of 7 to 15 dpa and containing 200 to 1000 appm He, a reduction in life by amore » factor of 3 to 10 was observed. An attempt was made to predict irradiated fatigue life by fitting data from irradiated material to a power law equation similar to the universal slopes equation and using ductility ratios from tensile tests to modify the equation for irradiated material.« less

High density operation for reactor-relevant power exhaust

NASA Astrophysics Data System (ADS)

Wischmeier, M.; ASDEX Upgrade Team; Jet Efda Contributors

2015-08-01

With increasing size of a tokamak device and associated fusion power gain an increasing power flux density towards the divertor needs to be handled. A solution for handling this power flux is crucial for a safe and economic operation. Using purely geometric arguments in an ITER-like divertor this power flux can be reduced by approximately a factor 100. Based on a conservative extrapolation of current technology for an integrated engineering approach to remove power deposited on plasma facing components a further reduction of the power flux density via volumetric processes in the plasma by up to a factor of 50 is required. Our current ability to interpret existing power exhaust scenarios using numerical transport codes is analyzed and an operational scenario as a potential solution for ITER like divertors under high density and highly radiating reactor-relevant conditions is presented. Alternative concepts for risk mitigation as well as strategies for moving forward are outlined.

HYFIRE II: fusion/high-temperature electrolysis conceptual-design study. Annual report

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fillo, J.A.

1983-08-01

As in the previous HYFIRE design study, the current study focuses on coupling a Tokamak fusion reactor with a high-temperature blanket to a High-Temperature Electrolyzer (HTE) process to produce hydrogen and oxygen. Scaling of the STARFIRE reactor to allow a blanket power to 6000 MW(th) is also assumed. The primary difference between the two studies is the maximum inlet steam temperature to the electrolyzer. This temperature is decreased from approx. 1300/sup 0/ to approx. 1150/sup 0/C, which is closer to the maximum projected temperature of the Westinghouse fuel cell design. The process flow conditions change but the basic design philosophymore » and approaches to process design remain the same as before. Westinghouse assisted in the study in the areas of systems design integration, plasma engineering, balance-of-plant design, and electrolyzer technology.« less

PROSPECT - A precision oscillation and spectrum experiment

NASA Astrophysics Data System (ADS)

Langford, T. J.; PROSPECT Collaboration

2015-08-01

Segmented antineutrino detectors placed near a compact research reactor provide an excellent opportunity to probe short-baseline neutrino oscillations and precisely measure the reactor antineutrino spectrum. Close proximity to a reactor combined with minimal overburden yield a high background environment that must be managed through shielding and detector technology. PROSPECT is a new experimental effort to detect reactor antineutrinos from the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory, managed by UT Battelle for the U.S. Department of Energy. The detector will use novel lithium-loaded liquid scintillator capable of neutron/gamma pulse shape discrimination and neutron capture tagging. These enhancements improve the ability to identify neutrino inverse-beta decays (IBD) and reject background events in analysis. Results from these efforts will be covered along with their implications for an oscillation search and a precision spectrum measurement.

Inherently Safe Fission Power System for Lunar Outposts

NASA Astrophysics Data System (ADS)

Schriener, Timothy M.; El-Genk, Mohamed S.

2013-09-01

This paper presents the Solid Core-Sectored Compact Reactor (SC-SCoRe) and power system for future lunar outposts. The power system nominally provides 38 kWe continuously for 21 years, employs static components and has no single point failures in reactor cooling or power generation. The reactor core has six sectors, each has a separate pair of primary and secondary loops with liquid NaK-56 working fluid, thermoelectric (TE) power conversion and heat-pipes radiator panels. The electromagnetic (EM) pumps in the primary and secondary loops, powered with separate TE power units, ensure operation reliability and passive decay heat removal from the reactor after shutdown. The reactor poses no radiological concerns during launch, and remains sufficiently subcritical, with the radial reflector dissembled, when submerged in wet sand and the core flooded with seawater, following a launch abort accident. After 300 years of storage below grade on the Moon, the total radioactivity in the post-operation reactor drops below 164 Ci, a low enough radioactivity for a recovery and safe handling of the reactor.

Heat exchanger for reactor core and the like

DOEpatents

Kaufman, Jay S.; Kissinger, John A.

1986-01-01

A compact bayonet tube type heat exchanger which finds particular application as an auxiliary heat exchanger for transfer of heat from a reactor gas coolant to a secondary fluid medium. The heat exchanger is supported within a vertical cavity in a reactor vessel intersected by a reactor coolant passage at its upper end and having a reactor coolant return duct spaced below the inlet passage. The heat exchanger includes a plurality of relatively short length bayonet type heat exchange tube assemblies adapted to pass a secondary fluid medium therethrough and supported by primary and secondary tube sheets which are releasibly supported in a manner to facilitate removal and inspection of the bayonet tube assemblies from an access area below the heat exchanger. Inner and outer shrouds extend circumferentially of the tube assemblies and cause the reactor coolant to flow downwardly internally of the shrouds over the tube bundle and exit through the lower end of the inner shroud for passage to the return duct in the reactor vessel.

Energetic particle-driven compressional Alfvén eigenmodes and prospects for ion cyclotron emission studies in fusion plasmas

NASA Astrophysics Data System (ADS)

Gorelenkov, N. N.

2016-10-01

As a fundamental plasma oscillation the compressional Alfvén waves (CAWs) are interesting for plasma scientists both academically and in applications for fusion plasmas. They are believed to be responsible for the ion cyclotron emission (ICE) observed in many tokamaks. The theory of CAW and ICE was significantly advanced at the end of 20th century in particular motivated by first DT experiments on TFTR and subsequent JET DT experimental studies. More recently, ICE theory was advanced by ST (or spherical torus) experiments with the detailed theoretical and experimental studies of the properties of each instability signal. There the instability responsible for ICE signals previously indistinguishable in high aspect ratio tokamaks became the subjects of experimental studies. We discuss further the prospects of ICE theory and its applications for future burning plasma experiments such as the ITER tokamak-reactor prototype being build in France where neutrons and gamma rays escaping the plasma create extremely challenging conditions for fusion alpha particle diagnostics. This manuscript has been authored by Princeton University under Contract Number DE-AC02-09CH11466 with the US Department of Energy. The United States Government retains and the publisher, by accepting the article for publication, acknowledges that the United States Government retains a non-exclusive, paid-up, irrevocable, world-wide license to publish or reproduce the published form of this manuscript, or allow others to do so, for United States Government purposes.

The external kink mode in diverted tokamaks

DOE PAGES

Turnbull, Alan D.; Hanson, Jeremy M.; Turco, Francesca; ...

2016-06-16

Here, an explanation is provided for the disruptive instability in diverted tokamaks when the safety factor at the 95% poloidal flux surface, q 95, is driven below 2.0. The instability is a resistive kink counterpart to the current-driven ideal mode that traditionally explained the corresponding disruption in limited cross-sections when q edge, the safety factor at the outermost closed flux surface, lies just below a rational value. Experimentally, external kink modes are observed in limiter configurations as the current in a tokamak is ramped up and q edge decreases through successive rational surfaces. For q edge < 2, the instabilitymore » is always encountered and is highly disruptive. However, diverted plasmas, in which q edge is formally infinite in the magnetohydrodynamic (MHD) model, have presented a longstanding difficulty since the theory would predict stability, yet, the disruptive limit occurs in practice when q 95, reaches 2. It is shown from numerical calculations that a resistive kink mode is linearly destabilized by the rapidly increasing resistivity at the plasma edge when q 95 < 2, but q edge >> 2. The resistive kink behaves much like the ideal kink with predominantly kink or interchange parity and no real sign of a tearing component. However, the growth rates scale with a fractional power of the resistivity near the q = 2 surface. The results have a direct bearing on the conventional edge cutoff procedures used in most ideal MHD codes, as well as implications for ITER and for future reactor options.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wang, Y.; Tobias, B.; Chang, Y. -T.

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. The microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These also have the potential to greatly advance microwavemore » fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfven eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today's most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.« less

Non-axisymmetric equilibrium reconstruction and suppression of density limit disruptions in a current-carrying stellarator

NASA Astrophysics Data System (ADS)

Ma, Xinxing; Ennis, D. A.; Hanson, J. D.; Hartwell, G. J.; Knowlton, S. F.; Maurer, D. A.

2017-10-01

Non-axisymmetric equilibrium reconstructions have been routinely performed with the V3FIT code in the Compact Toroidal Hybrid (CTH), a stellarator/tokamak hybrid. In addition to 50 external magnetic measurements, 160 SXR emissivity measurements are incorporated into V3FIT to reconstruct the magnetic flux surface geometry and infer the current distribution within the plasma. Improved reconstructions of current and q profiles provide insight into understanding the physics of density limit disruptions observed in current-carrying discharges in CTH. It is confirmed that the final scenario of the density limit of CTH plasmas is consistent with classic observations in tokamaks: current profile shrinkage leads to growing MHD instabilities (tearing modes) followed by a loss of MHD equilibrium. It is also observed that the density limit at a given current linearly increases with increasing amounts of 3D shaping fields. Consequently, plasmas with densities up to two times the Greenwald limit are attained. Equilibrium reconstructions show that addition of 3D fields effectively moves resonance surfaces towards the edge of the plasma where the current profile gradient is less, providing a stabilizing effect. This work is supported by US Department of Energy Grant No. DE-FG02-00ER54610.

Safety Testing of AGR-2 UCO Compacts 6-4-2 and 2-3-1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hunn, John D.; Morris, Robert N.; Baldwin, Charles A.

2017-08-01

Post-irradiation examination (PIE) and elevated-temperature safety testing are being performed on tristructural-isotropic (TRISO) coated-particle fuel compacts from the Advanced Gas Reactor (AGR) Fuel Development and Qualification Program second irradiation experiment (AGR-2). Details on this irradiation experiment have been previously reported [Collin 2014]. The AGR-2 PIE effort builds upon the understanding acquired throughout the AGR-1 PIE campaign [Demkowicz et al. 2015] and is establishing a database for the different AGR-2 fuel designs.

Results of high heat flux tests of tungsten divertor targets under plasma heat loads expected in ITER and tokamaks (review)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Budaev, V. P., E-mail: budaev@mail.ru

2016-12-15

Heat loads on the tungsten divertor targets in the ITER and the tokamak power reactors reach ~10MW m{sup −2} in the steady state of DT discharges, increasing to ~0.6–3.5 GW m{sup −2} under disruptions and ELMs. The results of high heat flux tests (HHFTs) of tungsten under such transient plasma heat loads are reviewed in the paper. The main attention is paid to description of the surface microstructure, recrystallization, and the morphology of the cracks on the target. Effects of melting, cracking of tungsten, drop erosion of the surface, and formation of corrugated and porous layers are observed. Production ofmore » submicron-sized tungsten dust and the effects of the inhomogeneous surface of tungsten on the plasma–wall interaction are discussed. In conclusion, the necessity of further HHFTs and investigations of the durability of tungsten under high pulsed plasma loads on the ITER divertor plates, including disruptions and ELMs, is stressed.« less

Simulation of the alpha particle heating and the helium ash source in an International Thermonuclear Experimental Reactor-like tokamak with an internal transport barrier

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ye, Lei, E-mail: lye@ipp.ac.cn; Guo, Wenfeng; Xiao, Xiaotao

2014-12-15

A guiding center orbit following code, which incorporates a set of non-singular coordinates for orbit integration, was developed and applied to investigate the alpha particle heating in an ITER-like tokamak with an internal transport barrier. It is found that a relatively large q (safety factor) value can significantly broaden the alpha heating profile in comparison with the local heating approximation; this broadening is due to the finite orbit width effects; when the orbit width is much smaller than the scale length of the alpha particle source profile, the heating profile agrees with the source profile, otherwise, the heating profile canmore » be significantly broadened. It is also found that the stagnation particles move to the magnetic axis during the slowing-down process, thus the effect of stagnation orbits is not beneficial to the helium ash removal. The source profile of helium ash is broadened in comparison with the alpha source profile, which is similar to the heating profile.« less

Tearing mode dynamics in the RFX-mod tokamak

NASA Astrophysics Data System (ADS)

Cordaro, Luigi; Zanca, Paolo; Zuin, Matteo; Auriemma, Fulvio; Martines, Emilio; Zaniol, Barbara; Pucella, Gianluca; Cavazzana, Roberto; de Masi, Gianluca; Fassina, Alessandro; Grenfell, Gustavo; Momo, Barbara; Spagnolo, Silvia; Spolaore, Monica; Vianello, Nicola

2017-10-01

The study of the physical mechanisms that influence the tearing mode (TM) rotation is of interest because, while in present day devices, a significant TM rotation can be induced by Neutral Beam Injection, future reactors, ITER included, are not expected to provide enough induced momentum. We present a study of tearing mode dynamics in the RFX-mod device, a Reserved Field Pinch in Padua (Italy) that can be run as low-current, circular tokamak. Magnetic, flow and kinetic measurements are integrated to characterize the (2,1) and (3,2) TMs fast rotation. We are especially interested to study the role played by the diamagnetic electron drift on the TM rotation, including the slowing down and the wall-locking phases. When the latter occurs, the radial magnetic field penetrates the shell and the TM amplitude increases at a rate given by the wall resistive time constant. This phenomenon can lead to a rapid discharge termination via a disruption. A comparison of experimental data with a two-fluid MHD cylindrical model has been used to interpret the observed TM fast rotation frequencies.

Is Helium-3 hype, hyperbole or a hopeful fuel for the future?

NASA Astrophysics Data System (ADS)

Mackowski, Maura J.

Sixty kilowatts of thermal power have been reached with deuterium/He-3 reaction on the JET reactor, and full scale study of the environmental impact of a tokamak D/He-3 reactor is now underway for NASA. He-3 is obtained from the decaying process undergone by tritium, but in nature, the source of He-3 is the sun. It is found in abundance in the lunar regolith. He-3 combines with deuterium in a fusion reaction generating very high amounts of energy, He-4 and protons. He-3 is economical; it could be moon mined and sold at a price comparable to oil. The energy released is roughly 70 percent efficient, and can be directly converted to electricity with solid-state converters. Also, reactors can be built cheaper, placed closer to cities, and maintained and decomissioned more easily than any other type of fission or fusion reactor, thus allowing faster commercialization and lower energy costs. He-3 is not very radioactive; however, the physics of its nuclear structure presents barriers to getting it to fuse. Other advantages of producing He-3 on the moon include obtaining gases necessary in moon colonies, and fuel for hydrogen rockets. Water, nickel and carbon-oxygen compounds can also be obtained that way.

Non-electric applications for magneto-inertial fusion

NASA Astrophysics Data System (ADS)

Slough, John

2016-10-01

In addition to the generation of commercial electric power, there are several other applications for an intense pulse of neutrons that would be produced by magneto-inertial fusion (MIF) systems. Many of these applications can be achieved without the need for a fully developed reactor at high gain, and could thus be pursued at a much earlier stage of development which would dramatically reduce the risk of the long-term development and concern for the expense of an all-encompassing, single use system such as the tokamak or stellerator. A short list of applications well suited for MIF would include: (1) production of radioisotopes for medical applications and research, (2) efficient, high power propulsion through direct fusion heating of lithium propellants (3) Noninvasive interrogation of objects for homeland security (4) neutron radiography and tomography (5) destruction of long-lived radioactive waste, and (6) breeding of proliferation proof fissile fuel for existing nuclear reactors. These applications could all be pursued at lower neutron yield, but clearly the energy goals are by far the most significant and far reaching such as applying fusion energy as a hybrid to enable thorium cycle reactors which produce very little waste compared to the current uranium reactors. A discussion of how MIF could be configured and utilized to realize several of these uses will be discussed.

Optimization of tritium breeding and shielding analysis to plasma in ITER fusion reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Indah Rosidah, M., E-mail: indah.maymunah@gmail.com; Suud, Zaki, E-mail: szaki@fi.itb.ac.id; Yazid, Putranto Ilham

The development of fusion energy is one of the important International energy strategies with the important milestone is ITER (International Thermonuclear Experimental Reactor) project, initiated by many countries, such as: America, Europe, and Japan who agreed to set up TOKAMAK type fusion reactor in France. In ideal fusion reactor the fuel is purely deuterium, but it need higher temperature of reactor. In ITER project the fuels are deuterium and tritium which need lower temperature of the reactor. In this study tritium for fusion reactor can be produced by using reaction of lithium with neutron in the blanket region. With themore » tritium breeding blanket which react between Li-6 in the blanket with neutron resulted from the plasma region. In this research the material used in each layer surrounding the plasma in the reactor is optimized. Moreover, achieving self-sufficiency condition in the reactor in order tritium has enough availability to be consumed for a long time. In order to optimize Tritium Breeding Ratio (TBR) value in the fusion reactor, there are several strategies considered here. The first requirement is making variation in Li-6 enrichment to be 60%, 70%, and 90%. But, the result of that condition can not reach TBR value better than with no enrichment. Because there is reduction of Li-7 percent when increasing Li-6 percent. The other way is converting neutron multiplier material with Pb. From this, we get TBR value better with the Be as neutron multiplier. Beside of TBR value, fusion reactor can analyze the distribution of neutron flux and dose rate of neutron to know the change of neutron concentration for each layer in reactor. From the simulation in this study, 97% neutron concentration can be absorbed by material in reactor, so it is good enough. In addition, it is required to analyze spectrum neutron energy in many layers in the fusion reactor such as in blanket, coolant, and divertor. Actually material in that layer can resist in high temperature and high pressure condition for more than ten years.« less

Kilowatt Reactor Using Stirling TechnologY (KRUSTY) Demonstration. CEDT Phase 1 Preliminary Design Documentation

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sanchez, Rene Gerardo; Hutchinson, Jesson D.; Mcclure, Patrick Ray

2015-08-20

The intent of the integral experiment request IER 299 (called KiloPower by NASA) is to assemble and evaluate the operational performance of a compact reactor configuration that closely resembles the flight unit to be used by NASA to execute a deep space exploration mission. The reactor design will include heat pipes coupled to Stirling engines to demonstrate how one can generate electricity when extracting energy from a “nuclear generated” heat source. This series of experiments is a larger scale follow up to the DUFF series of experiments1,2 that were performed using the Flat-Top assembly.

Theoretical transport modeling of Ohmic cold pulse experiments

NASA Astrophysics Data System (ADS)

Kinsey, J. E.; Waltz, R. E.; St. John, H. E.

1998-11-01

The response of several theory-based transport models in Ohmically heated tokamak discharges to rapid edge cooling due to trace impurity injection is studied. Results are presented for the Institute for Fusion Studies—Princeton Plasma Physics Laboratory (IFS/PPPL), gyro-Landau-fluid (GLF23), Multi-mode (MM), and the Itoh-Itoh-Fukuyama (IIF) transport models with an emphasis on results from the Texas Experimental Tokamak (TEXT) [K. W. Gentle, Nucl. Technol./Fusion 1, 479 (1981)]. It is found that critical gradient models containing a strong ion and electron temperature ratio dependence can exhibit behavior that is qualitatively consistent with experimental observation while depending solely on local parameters. The IFS/PPPL model yields the strongest response and demonstrates both rapid radial pulse propagation and a noticeable increase in the central electron temperature following a cold edge temperature pulse (amplitude reversal). Furthermore, the amplitude reversal effect is predicted to diminish with increasing electron density and auxiliary heating in agreement with experimental data. An Ohmic pulse heating effect due to rearrangement of the current profile is shown to contribute to the rise in the core electron temperature in TEXT, but not in the Joint European Tokamak (JET) [A. Tanga and the JET Team, in Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 65] and the Tokamak Fusion Test Reactor (TFTR) [R. J. Hawryluk, V. Arunsalam, M. G. Bell et al., in Plasma Physics and Controlled Nuclear Fusion Research 1986 (International Atomic Energy Agency, Vienna, 1987), Vol. 1, p. 51]. While this phenomenon is not necessarily a unique signature of a critical gradient, there is sufficient evidence suggesting that the apparent plasma response to edge cooling may not require any underlying nonlocal mechanism and may be explained within the context of the intrinsic properties of electrostatic drift wave-based models.

Lower Hybrid wave edge power loss quantification on the Alcator C-Mod tokamak

NASA Astrophysics Data System (ADS)

Faust, I. C.

2015-11-01

For the first time, the power deposition of Lower Hybrid RF waves into the edge plasma of a diverted tokamak has been systematically quantified. Edge deposition represents a parasitic loss of power that can greatly impact the use and efficiency of Lower Hybrid Current Drive (LHCD) at reactor-relevant densities. Through the use of a unique set of fast time resolution edge diagnostics, including innovative fast-thermocouples, an extensive set of Langmuir probes, and a Lyα ionization camera, the toroidal, poloidal and radial structure of the power deposition has been simultaneously determined. Power modulation was used to directly isolate the RF effects due to the prompt (t <τE) response of the scrape-off-layer (SOL) plasma to LHRF power. LHRF power was found to absorb more strongly in the edge at higher densities. It is found that a majority of this edge-deposited power is promptly conducted to the divertor. This correlates with the loss of current drive efficiency at high density previously observed on Alcator C-Mod, and displaying characteristics that contrast with the local RF edge absorption seen on other tokamaks. Measurements of ionization in the active divertor show dramatic changes due to LHRF power, implying that divertor region can be key for the LHRF edge power deposition physics. These observations support the existence a loss mechanism near the edge for LHRF at high density (ne > 1 . 0 .1020 [m-3]). Results will be shown addressing the distribution of power within the SOL, including the toroidal symmetry and radial distribution. These characteristics are important for deducing the cause of the reduced LHCD efficiency at high density and motivates the tailoring of wave propagation to minimize SOL interaction, for example, through the use of high-field-side launch. This work was performed on the Alcator C-Mod tokamak, a DoE Office of Science user facility, and is supported by USDoE award DE-FC02-99ER54512.

A New Equivalence Theory Method for Treating Doubly Heterogeneous Fuel - I. Theory

DOE PAGES

Williams, Mark L.; Lee, Deokjung; Choi, Sooyoung

2015-03-04

A new methodology has been developed to treat resonance self-shielding in doubly heterogeneous very high temperature gas-cooled reactor systems in which the fuel compact region of a reactor lattice consists of small fuel grains dispersed in a graphite matrix. This new method first homogenizes the fuel grain and matrix materials using an analytically derived disadvantage factor from a two-region problem with equivalence theory and intermediate resonance method. This disadvantage factor accounts for spatial self-shielding effects inside each grain within the framework of an infinite array of grains. Then the homogenized fuel compact is self-shielded using a Bondarenko method to accountmore » for interactions between the fuel compact regions in the fuel lattice. In the final form of the equations for actual implementations, the double-heterogeneity effects are accounted for by simply using a modified definition of a background cross section, which includes geometry parameters and cross sections for both the grain and fuel compact regions. With the new method, the doubly heterogeneous resonance self-shielding effect can be treated easily even with legacy codes programmed only for a singly heterogeneous system by simple modifications in the background cross section for resonance integral interpolations. This paper presents a detailed derivation of the new method and a sensitivity study of double-heterogeneity parameters introduced during the derivation. The implementation of the method and verification results for various test cases are presented in the companion paper.« less

Ion Bernstein wave heating research

NASA Astrophysics Data System (ADS)

Ono, Masayuki

1993-02-01

Ion Bernstein wave heating (IBWH) utilizes the ion Bernstein wave (IBW), a hot plasma wave, to carry the radio frequency (rf) power to heat the tokamak reactor core. Earlier wave accessibility studies have shown that this finite-Larmor-radius (FLR) mode should penetrate into a hot dense reactor plasma core without significant attenuation. Moreover, the IBW's low perpendicular phase velocity (ω/k⊥≊VTi≪Vα) greatly reduces the otherwise serious wave absorption by the 3.5 MeV fusion α particles. In addition, the property of IBW's that k⊥ρi≊1 makes localized bulk ion heating possible at the ion cyclotron harmonic layers. Such bulk ion heating can prove useful in optimizing fusion reactivity. In another vein, with proper selection of parameters, IBW's can be made subject to strong localized electron Landau damping near the major ion cyclotron harmonic resonance layers. This property can be useful, for example, for rf current drive in the reactor plasma core. IBW's can be excited with loop antennas or with a lower-hybrid-like waveguide launcher at the plasma edge, the latter structure being one that is especially compatible with reactor application. In either case, the mode at the plasma edge is an electron plasma wave (EPW). Deeper in the plasma, the EPW is mode transformed into an IBW. Such launching and mode transformation of IBW's were first demonstrated in experiments in the Advanced Concepts Torus-1 (ACT-1) [Phys. Rev. Lett. 45, 1105 (1980)] plasma torus and in particle simulation calculations. These and other aspects of IBW heating physics have been investigated through a number of experiments performed on ACT-1, the Japanese Institute of Plasma Physics Tokamak II-Upgrade (JIPPTII-U) [Phys. Rev. Lett. 54, 2339 (1985)], the Tokyo University Non-Circular Tokamak (TNT) [Nucl. Fusion 26, 1097 (1986)], the Princeton Large Tokamak (PLT) [Phys. Rev. Lett. 60, 294 (1988)], and Alcator-C [Phys. Rev. Lett. 60, 298 (1988)]. In these experiments both linear and nonlinear heating processes have been observed. Interestingly, improvement of plasma confinement was also observed in the PLT and Alcator-C experiments, opening up the possible use of IBW's for the active control of plasma transport. Two theoretical explanations have been proposed: one based on four-wave mixing of IBW with low-frequency turbulence, the other on the nonlinear generation of a velocity-shear layer. Both models are consistent with the observed threshold power level of a few hundred kW in the experiments. Experiments on lower field plasmas on JFTII-M [Eighth Topical Conference on Radio-Frequency Power in Plasmas, Irvine, CA, 1989 (American Institute of Physics, New York, 1989), p. 350] and DIII-D [Eighth Topical Conference on Radio-Frequency Power in Plasmas, Irvine, CA, 1989 (American Institute of Physics, New York, 1989), p. 314] have raised some concern with the IBW wave-launching process. The experiments showed serious impurity release from the walls but little or no core heating, a combination of circumstances strongly suggestive of edge heating. Possible parasitic channels could include the excitation of short wavelength modes by the Faraday shield's fringing fields, antenna-sheath-wave excitation, an axial-convective loss channel, and nonlinear processes such as parametric instability and ponderomotive effects. Suggested remedies include changes in the antenna phasing, the use of low-Z insulators, operating at higher frequencies, positioning the plasma differently with respect to the antenna, eliminating the Faraday shields, and using a waveguide launcher. The recent JIPPTII-U experiment, employing a 0-π phased antenna array with a higher frequency 130 MHz source, demonstrated that those remedies can indeed work. Looking to the future, one seeks additional ways in which IBWH can improve tokamak performance. The strong ponderomotive potential of the IBWH antenna may be used to stabilize external kinks and, acting as an rf limiter, to control the plasma edge. Control of the plasma pressure profile with local IBWH heating is already an important part of the Princeton Beta Experiment-Modified (PBX-M) [Ninth Topical Conference on Radio-Frequency Power in Plasmas, Charleston, SC, 1991 (American Institute of Physics, New York, 1991), p. 129] program in its exploration of the second-stability regime. Application of IBWH may also improve the performance of neutral beam heating and the efficiency and localization of lower-hybrid current drive for current profile control. Used with pellet injection, IBWH may also prolong the period of good confinement. The three planned high-power IBWH experiments covering vastly different parameters: f=40-80 MHz for PBX-M; f=130 MHz for JIPPT-II-U; and f=430 MHz for the Frascati Tokamak-Upgrade (FT-U) [16th European Physical Society Conference on Controlled Fusion and Plasma Physics, Venice, Italy, 1989 (European Physical Society, Amsterdam, 1989), Vol. III, p. 1069] appear to be well positioned to explore these possibilities and to clarify other issues including the physics of wave launching and associated nonlinear processes.

SP-100 flight qualification testing assessment

NASA Technical Reports Server (NTRS)

Jeanmougin, Nanette M.; Moore, Roger M.; Wait, David L.; Jacox, Michael G.

1988-01-01

The SP-100 is a compact space power system driven by a nuclear reactor that provides 100 kWe to the user at 200 VDC. The thermal energy generated by the nuclear reactor is converted into electrical energy by passive thermoelectric devices. Various options for tailoring the MIL-STD-1540B guidelines to the SP-100 nuclear power system are discussed. This study aids in selecting the appropriate qualification test program based on the cost, schedule, and test effectiveness of the various options.

NEUTRONIC REACTOR CONTROL ROD AND METHOD OF FABRICATION

DOEpatents

Porembka, S.W. Jr.

1961-06-27

A reactor control rod formed from a compacted powder dispersion is patented. The rod consists of titanium sheathed with a cladding alloy. The cladding alloy contains 1.3% to 1.6% by weight of tin, 0.07% to 0.12% by weight of chromium, 0.04% to 0.08% by weight of nickel, 0.09% to 0.16% by weight of iron, carbon not exceeding 0.05%, less than 0.5% by weight of incidental impurities, and the balance zirconium.

Development of an advanced antineutrino detector for reactor monitoring

DOE PAGES

Classen, T.; Bernstein, A.; Bowden, N. S.; ...

2014-11-05

We present the development of a compact antineutrino detector for the purpose of nuclear reactor monitoring, improving upon a previously successful design. Our paper will describe the design improvements of the detector which increases the antineutrino detection efficiency threefold over the previous effort. There are two main design improvements over previous generations of detectors for nuclear reactor monitoring: dual-ended optical readout and single volume detection mass. The dual-ended optical readout eliminates the need for fiducialization and increases the uniformity of the detector's optical response. The containment of the detection mass in a single active volume provides more target mass permore » detector footprint, a key design criteria for operating within a nuclear power plant. This technology could allow for real-time monitoring of the evolution of a nuclear reactor core, independent of reactor operator declarations of fuel inventories, and may be of interest to the safeguards community.« less

Steam reforming of heptane in a fluidized bed membrane reactor

NASA Astrophysics Data System (ADS)

Rakib, Mohammad A.; Grace, John R.; Lim, C. Jim; Elnashaie, Said S. E. H.

n-Heptane served as a model compound to study steam reforming of naphtha as an alternative feedstock to natural gas for production of pure hydrogen in a fluidized bed membrane reactor. Selective removal of hydrogen using Pd 77Ag 23 membrane panels shifted the equilibrium-limited reactions to greater conversion of the hydrocarbons and lower yields of methane, an intermediate product. Experiments were conducted with no membranes, with one membrane panel, and with six panels along the height of the reactor to understand the performance improvement due to hydrogen removal in a reactor where catalyst particles were fluidized. Results indicate that a fluidized bed membrane reactor (FBMR) can provide a compact reformer for pure hydrogen production from a liquid hydrocarbon feedstock at moderate temperatures (475-550 °C). Under the experimental conditions investigated, the maximum achieved yield of pure hydrogen was 14.7 moles of pure hydrogen per mole of heptane fed.

Effect of Catalytic Cylinders on Autothermal Reforming of Methane for Hydrogen Production in a Microchamber Reactor

PubMed Central

Yan, Yunfei; Guo, Hongliang; Zhang, Li; Zhu, Junchen; Yang, Zhongqing; Tang, Qiang; Ji, Xin

2014-01-01

A new multicylinder microchamber reactor is designed on autothermal reforming of methane for hydrogen production, and its performance and thermal behavior, that is, based on the reaction mechanism, is numerically investigated by varying the cylinder radius, cylinder spacing, and cylinder layout. The results show that larger cylinder radius can promote reforming reaction; the mass fraction of methane decreased from 26% to 21% with cylinder radius from 0.25 mm to 0.75 mm; compact cylinder spacing corresponds to more catalytic surface and the time to steady state is decreased from 40 s to 20 s; alteration of staggered and aligned cylinder layout at constant inlet flow rates does not result in significant difference in reactor performance and it can be neglected. The results provide an indication and optimize performance of reactor; it achieves higher conversion compared with other reforming reactors. PMID:25097877

Progress in space nuclear reactor power systems technology development - The SP-100 program

NASA Technical Reports Server (NTRS)

Davis, H. S.

1984-01-01

Activities related to the development of high-temperature compact nuclear reactors for space applications had reached a comparatively high level in the U.S. during the mid-1950s and 1960s, although only one U.S. nuclear reactor-powered spacecraft was actually launched. After 1973, very little effort was devoted to space nuclear reactor and propulsion systems. In February 1983, significant activities toward the development of the technology for space nuclear reactor power systems were resumed with the SP-100 Program. Specific SP-100 Program objectives are partly related to the determination of the potential performance limits for space nuclear power systems in 100-kWe and 1- to 100-MW electrical classes. Attention is given to potential missions and applications, regimes of possible space power applicability, safety considerations, conceptual system designs, the establishment of technical feasibility, nuclear technology, materials technology, and prospects for the future.

Lunar Resource Utilization: Development of a Reactor for Volatile Extraction from Regolith

NASA Technical Reports Server (NTRS)

Kleinhenz, Julie E.; Sacksteder, Kurt R.; Nayagam, Vedha

2007-01-01

The extraction and processing of planetary resources into useful products, known as In- Situ Resource Utilization (ISRU), will have a profound impact on the future of planetary exploration. One such effort is the RESOLVE (Regolith and Environment Science, Oxygen and Lunar Volatiles Extraction) Project, which aims to extract and quantify these resources. As part of the first Engineering Breadboard Unit, the Regolith Volatiles Characterization (RVC) reactor was designed and built at the NASA Glenn Research Center. By heating and agitating the lunar regolith, loosely bound volatiles, such as hydrogen and water, are released and stored in the reactor for later analysis and collection. Intended for operation on a robotic rover, the reactor features a lightweight, compact design, easy loading and unloading of the regolith, and uniform heating of the regolith by means of vibrofluidization. The reactor performance was demonstrated using regolith simulant, JSC1, with favorable results.

Development of a Radial Deconsolidation Method

DOE Office of Scientific and Technical Information (OSTI.GOV)

Helmreich, Grant W.; Montgomery, Fred C.; Hunn, John D.

2015-12-01

A series of experiments have been initiated to determine the retention or mobility of fission products* in AGR fuel compacts [Petti, et al. 2010]. This information is needed to refine fission product transport models. The AGR-3/4 irradiation test involved half-inch-long compacts that each contained twenty designed-to-fail (DTF) particles, with 20-μm thick carbon-coated kernels whose coatings were deliberately fabricated such that they would crack under irradiation, providing a known source of post-irradiation isotopes. The DTF particles in these compacts were axially distributed along the compact centerline so that the diffusion of fission products released from the DTF kernels would be radiallymore » symmetric [Hunn, et al. 2012; Hunn et al. 2011; Kercher, et al. 2011; Hunn, et al. 2007]. Compacts containing DTF particles were irradiated at Idaho National Laboratory (INL) at the Advanced Test Reactor (ATR) [Collin, 2015]. Analysis of the diffusion of these various post-irradiation isotopes through the compact requires a method to radially deconsolidate the compacts so that nested-annular volumes may be analyzed for post-irradiation isotope inventory in the compact matrix, TRISO outer pyrolytic carbon (OPyC), and DTF kernels. An effective radial deconsolidation method and apparatus appropriate to this application has been developed and parametrically characterized.« less

Review of the magnetic fusion program by the 1986 ERAB Fusion Panel

NASA Astrophysics Data System (ADS)

Davidson, Ronald C.

1987-09-01

The 1986 ERAB Fusion Panel finds that fusion energy continues to be an attractive energy source with great potential for the future, and that the magnetic fusion program continues to make substantial technical progress. In addition, fusion research advances plasma physics, a sophisticated and useful branch of applied science, as well as technologies important to industry and defense. These factors fully justify the substantial expenditures by the Department of Energy in fusion research and development (R&D). The Panel endorses the overall program direction, strategy, and plans, and recognizes the importance and timeliness of proceeding with a burning plasma experiment, such as the proposed Compact Ignition Tokamak (CIT) experiment.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, Thomas

The design features developed for the Spherical Tokamak (ST) in the PPPL pilot plant study was used as the starting point in developing designs to meet the mission of a Fusion Nuclear Science Facility (FNSF) considering a range of machine sizes based on the influence of tritium consumption and maintenance strategies. The compact nature of a steady state operated ST device for this mission pushes operating conditions and places challenges in the design of components, device maintenance and the integration of supports and services. This paper reviews the general arrangement, design details and maintenance strategy of the ST-FNSF device coremore » for a 1.6-m and 1.0-m device; operating points which bracket the region between purchasing and breeding tritium.« less

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

DOE PAGES

Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.; ...

2016-05-18

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.« less

Performance of AGR-1 high-temperature reactor fuel during post-irradiation heating tests

DOE Office of Scientific and Technical Information (OSTI.GOV)

Morris, Robert N.; Baldwin, Charles A.; Demkowicz, Paul A.

The fission product retention of irradiated low-enriched uranium oxide/uranium carbide tri-structural isotropic (TRISO) fuel compacts from the Advanced Gas-Cooled Reactor 1 (AGR-1) experiment has been evaluated at temperatures of 1600–1800 °C during post-irradiation safety tests. Fourteen compacts (a total of ~58,000 particles) with a burnup ranging from 13.4% to 19.1% fissions per initial metal atom (FIMA) have been tested using dedicated furnace systems at Idaho National Laboratory and Oak Ridge National Laboratory. The release of fission products 110mAg, 134Cs, 137Cs, 154Eu, 155Eu, 90Sr, and 85Kr was monitored while heating the fuel specimens in flowing helium. The behavior of silver, europium,more » and strontium appears to be dominated by inventory that was originally released through intact SiC coating layers during irradiation, but was retained in the compact at the end of irradiation and subsequently released during the safety tests. However, at a test temperature of 1800 °C, the data suggest that release of these elements through intact coatings may become significant after ~100 h. Cesium was very well retained by intact SiC layers, with a fractional release <5 × 10–6 after 300 h at 1600 °C or 100 h at 1800 °C. However, it was rapidly released from individual particles if the SiC layer failed, and therefore the overall cesium release fraction was dominated by the SiC defect and failure fractions in the fuel compacts. No complete TRISO coating layer failures were observed after 300 h at 1600 or 1700 °C, and 85Kr release was very low during the tests (particles with failed SiC, but intact outer pyrocarbon, retained most of their krypton). Krypton release from TRISO failures was only observed after ~210 h at 1800 °C in one compact. As a result, post-safety-test examination of fuel compacts and particles has focused on identifying specific particles from each compact with notable fission product release and detailed analysis of the coating layers to understand particle behavior.« less

Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Zhu, Y. L.; Xie, J. L., E-mail: jlxie@ustc.edu.cn; Yu, C. X.

2016-11-15

Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This “4th generation” MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven bymore » fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy “general optics structure” has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.« less

Millimeter-wave imaging diagnostics systems on the EAST tokamak (invited)

NASA Astrophysics Data System (ADS)

Zhu, Y. L.; Xie, J. L.; Yu, C. X.; Zhao, Z. L.; Gao, B. X.; Chen, D. X.; Liu, W. D.; Liao, W.; Qu, C. M.; Luo, C.; Hu, X.; Spear, A. G.; Luhmann, N. C.; Domier, C. W.; Chen, M.; Ren, X.; Tobias, B. J.

2016-11-01

Millimeter-wave imaging diagnostics, with large poloidal span and wide radial range, have been developed on the EAST tokamak for visualization of 2D electron temperature and density fluctuations. A 384 channel (24 poloidal × 16 radial) Electron Cyclotron Emission Imaging (ECEI) system in F-band (90-140 GHz) was installed on the EAST tokamak in 2012 to provide 2D electron temperature fluctuation images with high spatial and temporal resolution. A co-located Microwave Imaging Reflectometry (MIR) will be installed for imaging of density fluctuations by December 2016. This "4th generation" MIR system has eight independent frequency illumination beams in W-band (75-110 GHz) driven by fast tuning synthesizers and active multipliers. Both of these advanced millimeter-wave imaging diagnostic systems have applied the latest techniques. A novel design philosophy "general optics structure" has been employed for the design of the ECEI and MIR receiver optics with large aperture. The extended radial and poloidal coverage of ECEI on EAST is made possible by innovations in the design of front-end optics. The front-end optical structures of the two imaging diagnostics, ECEI and MIR, have been integrated into a compact system, including the ECEI receiver and MIR transmitter and receiver. Two imaging systems share the same mid-plane port for simultaneous, co-located 2D fluctuation measurements of electron density and temperature. An intelligent remote-control is utilized in the MIR electronics systems to maintain focusing at the desired radial region even with density variations by remotely tuning the probe frequencies in about 200 μs. A similar intelligent technique has also been applied on the ECEI IF system, with remote configuration of the attenuations for each channel.

Non-Nuclear Testing of Compact Reactor Technologies at NASA MSFC

NASA Technical Reports Server (NTRS)

Houts, Michael G.; Pearson, J. Boise; Godfroy, Thomas J.

2011-01-01

Safe, reliable, compact, autonomous, long-life fission systems have numerous potential applications, both terrestrially and in space. Technologies and facilities developed in support of these systems could be useful to a variety of concepts. At moderate power levels, fission systems can be designed to operate for decades without the need for refueling. In addition, fast neutron damage to cladding and structural materials can be maintained at an acceptable level. Nuclear design codes have advanced to the stage where high confidence in the behavior and performance of a system can be achieved prior to initial testing. To help ensure reactor affordability, an optimal strategy must be devised for development and qualification. That strategy typically involves a combination of non-nuclear and nuclear testing. Non-nuclear testing is particularly useful for concepts in which nuclear operating characteristics are well understood and nuclear effects such as burnup and radiation damage are not likely to be significant. To be mass efficient, a SFPS must operate at higher coolant temperatures and use different types of power conversion than typical terrestrial reactors. The primary reason is the difficulty in rejecting excess heat to space. Although many options exist, NASA s current reference SFPS uses a fast spectrum, pumped-NaK cooled reactor coupled to a Stirling power conversion subsystem. The reference system uses technology with significant terrestrial heritage while still providing excellent performance. In addition, technologies from the SFPS system could be applicable to compact terrestrial systems. Recent non-nuclear testing at NASA s Early Flight Fission Test Facility (EFF-TF) has helped assess the viability of the reference SFPS and evaluate methods for system integration. In July, 2011 an Annular Linear Induction Pump (ALIP) provided by Idaho National Laboratory was tested at the EFF-TF to assess performance and verify suitability for use in a10 kWe technology demonstration unit (TDU). In November, 2011 testing of a 37-pin core simulator (designed in conjunction with Los Alamos National Laboratory) for use with the TDU will occur. Previous testing at the EFFTF has included the thermal and mechanical coupling of a pumped NaK loop to Stirling engines (provided by GRC). Testing related to heat pipe cooled systems, gas cooled systems, heat exchangers, and other technologies has also been performed. Integrated TDU testing will begin at GRC in 2013. Thermal simulators developed at the EFF-TF are capable of operating over the temperature and power range typically of interest to compact reactors. Small and large diameter simulators have been developed, and simulators (coupled with the facility) are able to closely match the axial and radial power profile of all potential systems of interest. A photograph of the TDU core simulator during assembly is provided in Figure 2.

Compaction in optical fibres and fibre Bragg gratings under nuclear reactor high neutron and gamma fluence

DOE Office of Scientific and Technical Information (OSTI.GOV)

Remy, L.; Cheymol, G.; Gusarov, A.

2015-07-01

In the framework of the development by CEA and SCK.CEN of a Fabry Perot Sensor (FPS) able to measure dimensional changes in Material Testing Reactor (MTR), the first goal of the SAKE 1 (Smirnof extention - Additional Key-tests on Elongation of glass fibres) irradiation was to measure the linear compaction of single mode fibres under high fast neutron fluence. Indeed, the compaction of the fibre which forms one side of the Fabry Perot cavity, may in particular cause a noticeable measurement error. An accurate quantification of this effect is then required to predict the radiation-induced drift and optimize the sensormore » design. To achieve this, an innovative approach was used. Approximately seventy uncoated fibre tips (length: 30 to 50 mm) have been prepared from several different fibre samples and were installed in the SCK.CEN BR2 reactor (Mol Belgium). After 22 days of irradiation a total fast (E > 1 MeV) fluence of 3 to 5x10{sup 19} n{sub fast}/cm{sup 2}, depending on the sample location, was accumulated. The temperature during irradiation was 291 deg. C, which is not far from the condition of the intended FPS use. A precise measurement of each fibre tip length was made before the irradiation and compared to the post irradiation measurement highlighting a decrease of the fibres' length corresponding to about 0.25% of linear compaction. The amplitude of the changes is independent of the capsule, which could mean that the compaction effect saturates even at the lowest considered fluence. In the prospect of performing distributed temperature measurement in MTR, several fibre Bragg gratings written using a femtosecond laser have been also irradiated. All the gratings were written in radiation hardened fibres, and underwent an additional treatment with a procedure enhancing their resistance to ionizing radiations. A special mounting made it possible to test the reflection and the transmission of the gratings on fibre samples cut down to 30 to 50 mm. The comparison of measurements made before and after the irradiation, at the same temperature, allowed us to measure the loss in reflectivity as well as the Bragg wavelength drift. The results are quite promising for some of the investigated gratings. (authors)« less

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

DOE PAGES

Frattolillo, Antonio; Baylor, Larry R.; Bombarda, Francesca; ...

2018-04-17

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in themore » 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.« less

R&D around a photoneutralizer-based NBI system (Siphore) in view of a DEMO Tokamak steady state fusion reactor

NASA Astrophysics Data System (ADS)

Simonin, A.; Achard, Jocelyn; Achkasov, K.; Bechu, S.; Baudouin, C.; Baulaigue, O.; Blondel, C.; Boeuf, J. P.; Bresteau, D.; Cartry, G.; Chaibi, W.; Drag, C.; de Esch, H. P. L.; Fiorucci, D.; Fubiani, G.; Furno, I.; Futtersack, R.; Garibaldi, P.; Gicquel, A.; Grand, C.; Guittienne, Ph.; Hagelaar, G.; Howling, A.; Jacquier, R.; Kirkpatrick, M. J.; Lemoine, D.; Lepetit, B.; Minea, T.; Odic, E.; Revel, A.; Soliman, B. A.; Teste, P.

2015-11-01

Since the signature of the ITER treaty in 2006, a new research programme targeting the emergence of a new generation of neutral beam (NB) system for the future fusion reactor (DEMO Tokamak) has been underway between several laboratories in Europe. The specifications required to operate a NB system on DEMO are very demanding: the system has to provide plasma heating, current drive and plasma control at a very high level of power (up to 150 MW) and energy (1 or 2 MeV), including high performances in term of wall-plug efficiency (η  >  60%), high availability and reliability. To this aim, a novel NB concept based on the photodetachment of the energetic negative ion beam is under study. The keystone of this new concept is the achievement of a photoneutralizer where a high power photon flux (~3 MW) generated within a Fabry-Perot cavity will overlap, cross and partially photodetach the intense negative ion beam accelerated at high energy (1 or 2 MeV). The aspect ratio of the beam-line (source, accelerator, etc) is specifically designed to maximize the overlap of the photon beam with the ion beam. It is shown that such a photoneutralized based NB system would have the capability to provide several tens of MW of D0 per beam line with a wall-plug efficiency higher than 60%. A feasibility study of the concept has been launched between different laboratories to address the different physics aspects, i.e. negative ion source, plasma modelling, ion accelerator simulation, photoneutralization and high voltage holding under vacuum. The paper describes the present status of the project and the main achievements of the developments in laboratories.

Phase 3 experiments of the JAERI/USDOE collaborative program on fusion blanket neutronics. Volume 1: Experiment

NASA Astrophysics Data System (ADS)

Oyama, Yukio; Konno, Chikara; Ikeda, Yujiro; Maekawa, Fujio; Kosako, Kazuaki; Nakamura, Tomoo; Maekawa, Hiroshi; Youssef, Mahmoud Z.; Kumar, Anil; Abdou, Mohamed A.

1994-02-01

A pseudo-line source has been realized by using an accelerator based D-T point neutron source. The pseudo-line source is obtained by time averaging of continuously moving point source or by superposition of finely distributed point sources. The line source is utilized for fusion blanket neutronics experiments with an annular geometry so as to simulate a part of a tokamak reactor. The source neutron characteristics were measured for two operational modes for the line source, continuous and step-wide modes, with the activation foil and the NE213 detectors, respectively. In order to give a source condition for a successive calculational analysis on the annular blanket experiment, the neutron source characteristics was calculated by a Monte Carlo code. The reliability of the Monte Carlo calculation was confirmed by comparison with the measured source characteristics. The shape of the annular blanket system was a rectangular with an inner cavity. The annular blanket was consist of 15 mm-thick first wall (SS304) and 406 mm-thick breeder zone with Li2O at inside and Li2CO3 at outside. The line source was produced at the center of the inner cavity by moving the annular blanket system in the span of 2 m. Three annular blanket configurations were examined; the reference blanket, the blanket covered with 25 mm thick graphite armor and the armor-blanket with a large opening. The neutronics parameters of tritium production rate, neutron spectrum and activation reaction rate were measured with specially developed techniques such as multi-detector data acquisition system, spectrum weighting function method and ramp controlled high voltage system. The present experiment provides unique data for a higher step of benchmark to test a reliability of neutronics design calculation for a realistic tokamak reactor.

Whistlers, Helicons, Lower Hybrid Waves: the Physics of RF Wave Absorption Without Cyclotron Resonances

NASA Astrophysics Data System (ADS)

Pinsker, R. I.

2014-10-01

In hot magnetized plasmas, two types of linear collisionless absorption processes are used to heat and drive noninductive current: absorption at ion or electron cyclotron resonances and their harmonics, and absorption by Landau damping and the transit-time-magnetic-pumping (TTMP) interactions. This tutorial discusses the latter process, i.e., parallel interactions between rf waves and electrons in which cyclotron resonance is not involved. Electron damping by the parallel interactions can be important in the ICRF, particularly in the higher harmonic region where competing ion cyclotron damping is weak, as well as in the Lower Hybrid Range of Frequencies (LHRF), which is in the neighborhood of the geometric mean of the ion and electron cyclotron frequencies. On the other hand, absorption by parallel processes is not significant in conventional ECRF schemes. Parallel interactions are especially important for the realization of high current drive efficiency with rf waves, and an application of particular recent interest is current drive with the whistler or helicon wave at high to very high (i.e., the LHRF) ion cyclotron harmonics. The scaling of absorption by parallel interactions with wave frequency is examined and the advantages and disadvantages of fast (helicons/whistlers) and slow (lower hybrid) waves in the LHRF in the context of reactor-grade tokamak plasmas are compared. In this frequency range, both wave modes can propagate in a significant fraction of the discharge volume; the ways in which the two waves can interact with each other are considered. The use of parallel interactions to heat and drive current in practice will be illustrated with examples from past experiments; also looking forward, this tutorial will provide an overview of potential applications in tokamak reactors. Supported by the US Department of Energy under DE-FC02-04ER54698.

Core Fueling of DEMO by Direct Line Injection of High-Speed Pellets From the HFS

DOE Office of Scientific and Technical Information (OSTI.GOV)

Frattolillo, Antonio; Baylor, Larry R.; Bombarda, Francesca

Pellet injection represents to date the most realistic candidate technology for core fueling of a demonstration fusion power reactor tokamak fusion reactor. Modeling of both pellet penetration and fuel deposition profiles, for different injection locations, indicates that effective core fuelling can be achieved launching pellets from the inboard high field side at speeds not less than ~ 1 km/s. Inboard pellet fueling is commonly achieved in present tokamaks, using curved guide tubes; however, this technology might be hampered at velocities ≥ 1 km/s. An innovative approach, aimed at identifying suitable inboard "direct line'' paths, to inject high-speed pellets (in themore » 3 to 4 km/s range), has recently been proposed as a potential complementary solution. The fuel deposition profiles achievable by this approach have been explored using the HPI2 simulation code. The results presented here show that there are possible geometrical schemes providing good fueling performance. The problem of neutron flux in a direct line-of-sight injection path is being investigated, though preliminary analyses indicate that, perhaps, this is not a serious problem. The identification and integration of straight injection paths suitably tilted may be a rather difficult task due to the many constraints and to interference with existing structures. The suitability of straight guide tubes to reduce the scatter cone of high-speed pellets is, therefore, of main interest. A preliminary investigation, aimed at addressing these technological issues, has recently been started. As a result, a possible implementation plan, using an existing Italian National Agency for New Technologies, Energy and Sustainable Economic Development-Oak Ridge National Laboratory facility is shortly outlined.« less

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

NASA Astrophysics Data System (ADS)

Hamann, S.; Börner, K.; Burlacov, I.; Spies, H.-J.; Strämke, M.; Strämke, S.; Röpcke, J.

2015-12-01

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steel samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH4, C2H2, HCN, and NH3). With the help of OES, the rotational temperature of the screen plasma could be determined.

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-04-04

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Spherical torus fusion reactor

DOEpatents

Peng, Yueng-Kay M.

1989-01-01

A fusion reactor is provided having a near spherical-shaped plasma with a modest central opening through which straight segments of toroidal field coils extend that carry electrical current for generating a toroidal magnet plasma confinement fields. By retaining only the indispensable components inboard of the plasma torus, principally the cooled toroidal field conductors and in some cases a vacuum containment vessel wall, the fusion reactor features an exceptionally small aspect ratio (typically about 1.5), a naturally elongated plasma cross section without extensive field shaping, requires low strength magnetic containment fields, small size and high beta. These features combine to produce a spherical torus plasma in a unique physics regime which permits compact fusion at low field and modest cost.

Compact advanced extreme-ultraviolet imaging spectrometer for spatiotemporally varying tungsten spectra from fusion plasmas.

PubMed

Song, Inwoo; Seon, C R; Hong, Joohwan; An, Y H; Barnsley, R; Guirlet, R; Choe, Wonho

2017-09-01

A compact advanced extreme-ultraviolet (EUV) spectrometer operating in the EUV wavelength range of a few nanometers to measure spatially resolved line emissions from tungsten (W) was developed for studying W transport in fusion plasmas. This system consists of two perpendicularly crossed slits-an entrance aperture and a space-resolved slit-inside a chamber operating as a pinhole, which enables the system to obtain a spatial distribution of line emissions. Moreover, a so-called v-shaped slit was devised to manage the aperture size for measuring the spatial resolution of the system caused by the finite width of the pinhole. A back-illuminated charge-coupled device was used as a detector with 2048 × 512 active pixels, each with dimensions of 13.5 × 13.5 μm 2 . After the alignment and installation on Korea superconducting tokamak advanced research, the preliminary results were obtained during the 2016 campaign. Several well-known carbon atomic lines in the 2-7 nm range originating from intrinsic carbon impurities were observed and used for wavelength calibration. Further, the time behavior of their spatial distributions is presented.

Ball-and-Socket-Bearing Wear Test

NASA Technical Reports Server (NTRS)

Graham, W. G.

1984-01-01

Series of experiments to measure wear life of spherical bearing summarized. Report designed to establish clearance, contour, finish, and lubricant parameters for highly-loaded, compact plain spherical bearing. Information useful in design of bearings for helicopter control linkages, business machines, nuclear reactor, and rotor bearings.

STUDIES ON CONTAMINANT BIODEGRADATION IN SLURRY, WAFER, AND COMPACTED SOIL TUBE REACTORS

EPA Science Inventory

A systematic experimental approach is presented to quantitatively evaluate biodegradation rates in intact soil systems. Knowledge of bioremediation rates in intact soil systems is important for evaluating the efficacy of in-situ biodegradation and approaches for enhancing degrad...

Feasibility of recycling thorium in a fusion-fission hybrid/PWR symbiotic system

DOE Office of Scientific and Technical Information (OSTI.GOV)

Josephs, J.M.

1980-12-31

A study was made of the economic impact of high levels of radioactivity in the thorium fuel cycle. The sources of this radioactivity and means of calculating the radioactive levels at various stages in the fuel cycle are discussed and estimates of expected levels are given. The feasibility of various methods of recycling thorium is discussed. These methods include direct recycle, recycle after storage for 14 years to allow radioactivity to decrease, shortening irradiation times to limit radioactivity build up, and the use of the window in time immediately after reprocessing where radioactivity levels are diminished. An economic comparison ismore » made for the first two methods together with the throwaway option where thorium is not recycled using a mass energy flow model developed for a CTHR (Commercial Tokamak Hybrid Reactor), a fusion fission hybrid reactor which serves as fuel producer for several PWR reactors. The storage option is found to be most favorable; however, even this option represents a significant economic impact due to radioactivity of 0.074 mills/kW-h which amounts to $4 x 10/sup 9/ over a 30 year period assuming a 200 gigawatt supply of electrical power.« less

Demonstrating electromagnetic control of free-surface, liquid-metal flows relevant to fusion reactors

NASA Astrophysics Data System (ADS)

Hvasta, M. G.; Kolemen, E.; Fisher, A. E.; Ji, H.

2018-01-01

Plasma-facing components (PFC’s) made from solid materials may not be able to withstand the large heat and particle fluxes that will be produced within next-generation fusion reactors. To address the shortcomings of solid PFC’s, a variety of liquid-metal (LM) PFC concepts have been proposed. Many of the suggested LM-PFC designs rely on electromagnetic restraint (Lorentz force) to keep free-surface, liquid-metal flows adhered to the interior surfaces of a fusion reactor. However, there is very little, if any, experimental data demonstrating that free-surface, LM-PFC’s can actually be electromagnetically controlled. Therefore, in this study, electrical currents were injected into a free-surface liquid-metal that was flowing through a uniform magnetic field. The resultant Lorentz force generated within the liquid-metal affected the velocity and depth of the flow in a controllable manner that closely matched theoretical predictions. These results show the promise of electromagnetic control for LM-PFC’s and suggest that electromagnetic control could be further developed to adjust liquid-metal nozzle output, prevent splashing within a tokamak, and alter heat transfer properties for a wide-range of liquid-metal systems.

Demonstrating electromagnetic control of free-surface, liquid-metal flows relevant to fusion reactors

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hvasta, Michael George; Kolemen, Egemen; Fisher, Adam

Plasma-facing components (PFC's) made from solid materials may not be able to withstand the large heat and particle fluxes that will be produced within next-generation fusion reactors. To address the shortcomings of solid PFC's, a variety of liquid-metal (LM) PFC concepts have been proposed. Many of the suggested LM-PFC designs rely on electromagnetic restraint (Lorentz force) to keep free-surface, liquid-metal flows adhered to the interior surfaces of a fusion reactor. However, there is very little, if any, experimental data demonstrating that free-surface, LM-PFC's can actually be electromagnetically controlled. Therefore, in this study, electrical currents were injected into a free-surface liquid-metalmore » that was flowing through a uniform magnetic field. The resultant Lorentz force generated within the liquid-metal affected the velocity and depth of the flow in a controllable manner that closely matched theoretical predictions. Furthermore, these results show the promise of electromagnetic control for LM-PFC's and suggest that electromagnetic control could be further developed to adjust liquid-metal nozzle output, prevent splashing within a tokamak, and alter heat transfer properties for a wide-range of liquid-metal systems.« less

Demonstrating electromagnetic control of free-surface, liquid-metal flows relevant to fusion reactors

DOE PAGES

Hvasta, Michael George; Kolemen, Egemen; Fisher, Adam; ...

2017-10-13

Plasma-facing components (PFC's) made from solid materials may not be able to withstand the large heat and particle fluxes that will be produced within next-generation fusion reactors. To address the shortcomings of solid PFC's, a variety of liquid-metal (LM) PFC concepts have been proposed. Many of the suggested LM-PFC designs rely on electromagnetic restraint (Lorentz force) to keep free-surface, liquid-metal flows adhered to the interior surfaces of a fusion reactor. However, there is very little, if any, experimental data demonstrating that free-surface, LM-PFC's can actually be electromagnetically controlled. Therefore, in this study, electrical currents were injected into a free-surface liquid-metalmore » that was flowing through a uniform magnetic field. The resultant Lorentz force generated within the liquid-metal affected the velocity and depth of the flow in a controllable manner that closely matched theoretical predictions. Furthermore, these results show the promise of electromagnetic control for LM-PFC's and suggest that electromagnetic control could be further developed to adjust liquid-metal nozzle output, prevent splashing within a tokamak, and alter heat transfer properties for a wide-range of liquid-metal systems.« less

Developing snowflake divertor physics basis in the DIII-D, NSTX and NSTX-U tokamaks aimed at the divertor power exhaust solution [Snowflake divertor experiments in the DIII-D, NSTX and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-06-02

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment were performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large ELMs. However, a stable partial detachment of the outer strike pointmore » was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (cf. standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower n e, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at P SOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multi-fluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected P SOL ≃9 MW case. In conclusion, the radiative SF divertor with carbon impurity provides a wider n e operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (cf. standard divertor).« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finley, V.L.; Wiezcorek, M.A.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory (PPPL) for CY93. The report is prepared to provide the U.S. Department of Energy (DOE) and the public with information on the level of radioactive and non-radioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs that were undertaken in 1993. The objective of the Annual Site Environmental Report is to document evidence that DOE facility environmental protection programs adequately protect the environment and the public health. The Princeton Plasmamore » Physics Laboratory has engaged in fusion energy research since 1951. The long-range goal of the U.S. Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. In 1993, PPPL had both of its two large tokamak devices in operation; the Tokamak Fusion Test Reactor (TFTR) and the Princeton Beta Experiment-Modification (PBX-M). PBX-M completed its modifications and upgrades and resumed operation in November 1991. TFTR began the deuterium-tritium (D-T) experiments in December 1993 and set new records by producing over six million watts of energy. The engineering design phase of the Tokamak Physics Experiment (TPX), which replaced the cancelled Burning Plasma Experiment in 1992 as PPPL`s next machine, began in 1993 with the planned start up set for the year 2001. In 1993, the Environmental Assessment (EA) for the TFRR Shutdown and Removal (S&R) and TPX was prepared for submittal to the regulatory agencies.« less

Steady RANS methodology for calculating pressure drop in an in-line molten salt compact crossflow heat exchanger

DOE PAGES

Carasik, Lane B.; Shaver, Dillon R.; Haefner, Jonah B.; ...

2017-08-21

We report the development of molten salt cooled reactors (MSR) and fluoride-salt cooled high temperature reactors (FHR) requires the use of advanced design tools for the primary heat exchanger design. Due to geometric and flow characteristics, compact (pitch to diameter ratios equal to or less than 1.25) heat exchangers with a crossflow flow arrangement can become desirable for these reactors. Unfortunately, the available experimental data is limited for compact tube bundles or banks in crossflow. Computational Fluid Dynamics can be used to alleviate the lack of experimental data in these tube banks. Previous computational efforts have been primarily focused onmore » large S/D ratios (larger than 1.4) using unsteady Reynolds averaged Navier-Stokes and Large Eddy Simulation frameworks. These approaches are useful, but have large computational requirements that make comprehensive design studies impractical. A CFD study was conducted with steady RANS in an effort to provide a starting point for future design work. The study was performed for an in-line tube bank geometry with FLiBe (LiF-BeF2), a frequently selected molten salt, as the working fluid. Based on the estimated pressure drops, the pressure and velocity distributions in the domain, an appropriate meshing strategy was determined and presented. Periodic boundaries in the spanwise direction transverse flow were determined to be an appropriate boundary condition for reduced computational domains. The domain size was investigated and a minimum of 2-flow channels for a domain is recommended to ensure the behavior is accounted for. Finally, the standard low Re κ-ε (Lien) turbulence model was determined to be the most appropriate for steady RANS of this case at the time of writing.« less

Steady RANS methodology for calculating pressure drop in an in-line molten salt compact crossflow heat exchanger

DOE Office of Scientific and Technical Information (OSTI.GOV)

Carasik, Lane B.; Shaver, Dillon R.; Haefner, Jonah B.

We report the development of molten salt cooled reactors (MSR) and fluoride-salt cooled high temperature reactors (FHR) requires the use of advanced design tools for the primary heat exchanger design. Due to geometric and flow characteristics, compact (pitch to diameter ratios equal to or less than 1.25) heat exchangers with a crossflow flow arrangement can become desirable for these reactors. Unfortunately, the available experimental data is limited for compact tube bundles or banks in crossflow. Computational Fluid Dynamics can be used to alleviate the lack of experimental data in these tube banks. Previous computational efforts have been primarily focused onmore » large S/D ratios (larger than 1.4) using unsteady Reynolds averaged Navier-Stokes and Large Eddy Simulation frameworks. These approaches are useful, but have large computational requirements that make comprehensive design studies impractical. A CFD study was conducted with steady RANS in an effort to provide a starting point for future design work. The study was performed for an in-line tube bank geometry with FLiBe (LiF-BeF2), a frequently selected molten salt, as the working fluid. Based on the estimated pressure drops, the pressure and velocity distributions in the domain, an appropriate meshing strategy was determined and presented. Periodic boundaries in the spanwise direction transverse flow were determined to be an appropriate boundary condition for reduced computational domains. The domain size was investigated and a minimum of 2-flow channels for a domain is recommended to ensure the behavior is accounted for. Finally, the standard low Re κ-ε (Lien) turbulence model was determined to be the most appropriate for steady RANS of this case at the time of writing.« less

Overview of Indian activities on fusion reactor materials

NASA Astrophysics Data System (ADS)

Banerjee, Srikumar

2014-12-01

This paper on overview of Indian activities on fusion reactor materials describes in brief the efforts India has made to develop materials for the first wall of a tokamak, its blanket and superconducting magnet coils. Through a systematic and scientific approach, India has developed and commercially produced reduced activation ferritic/martensitic (RAFM) steel that is comparable to Eurofer 97. Powder of low activation ferritic/martensitic oxide dispersion strengthened steel with characteristics desired for its application in the first wall of a tokamak has been produced on the laboratory scale. V-4Cr-4Ti alloy was also prepared in the laboratory, and kinetics of hydrogen absorption in this was investigated. Cu-1 wt%Cr-0.1 wt%Zr - an alloy meant for use as heat transfer elements for hypervapotrons and heat sink for the first wall - was developed and characterized in detail for its aging behavior. The role of addition of a small quantity of Zr in its improved fatigue performance was delineated, and its diffusion bonding with both W and stainless steel was achieved using Ni as an interlayer. The alloy was produced in large quantities and used for manufacturing both the heat transfer elements and components for the International Thermonuclear Experimental Reactor (ITER). India has proposed to install and test a lead-lithium cooled ceramic breeder test blanket module (LLCB-TBM) at ITER. To meet this objective, efforts have been made to produce and characterize Li2TiO3 pebbles, and also improve the thermal conductivity of packed beds of these pebbles. Liquid metal loops have been set up and corrosion behavior of RAFM steel in flowing Pb-Li eutectic has been studied in the presence as well as absence of magnetic fields. To prevent permeation of tritium and reduce the magneto-hydro-dynamic drag, processes have been developed for coating alumina on RAFM steel. Apart from these activities, different approaches being attempted to make the U-shaped first wall of the TBM box are briefly described. India has also initiated the development of fusion grade superconductors. Success achieved in the fabrication of Nb3Sn based multi-filamentary wires using the internal tin process and cable-in-conduit-conductors is also briefly presented.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Venkataraman, M.; Natarajan, R.; Raj, Baldev

The reprocessing of spent fuel from Fast Breeder Test Reactor (FBTR) has been successfully demonstrated in the pilot plant, CORAL (COmpact Reprocessing facility for Advanced fuels in Lead shielded cell). Since commissioning in 2003, spent mixed carbide fuel from FBTR of different burnups and varying cooling period, have been reprocessed in this facility. Reprocessing of the spent fuel with a maximum burnup of 100 GWd/t has been successfully carried out so far. The feed backs from these campaigns with progressively increasing specific activities, have been useful in establishing a viable process flowsheet for reprocessing the Prototype Fast Breeder Reactor (PFBR)more » spent fuel. Also, the design of various equipments and processes for the future plants, which are either under design for construction, namely, the Demonstration Fast Reactor Fuel Reprocessing Plant (DFRP) and the Fast reactor fuel Reprocessing Plant (FRP) could be finalized. (authors)« less

Conceptual design of fast-ignition laser fusion reactor FALCON-D

NASA Astrophysics Data System (ADS)

Goto, T.; Someya, Y.; Ogawa, Y.; Hiwatari, R.; Asaoka, Y.; Okano, K.; Sunahara, A.; Johzaki, T.

2009-07-01

A new conceptual design of the laser fusion power plant FALCON-D (Fast-ignition Advanced Laser fusion reactor CONcept with a Dry wall chamber) has been proposed. The fast-ignition method can achieve sufficient fusion gain for a commercial operation (~100) with about 10 times smaller fusion yield than the conventional central ignition method. FALCON-D makes full use of this property and aims at designing with a compact dry wall chamber (5-6 m radius). 1D/2D simulations by hydrodynamic codes showed a possibility of achieving sufficient gain with a laser energy of 400 kJ, i.e. a 40 MJ target yield. The design feasibility of the compact dry wall chamber and the solid breeder blanket system was shown through thermomechanical analysis of the dry wall and neutronics analysis of the blanket system. Moderate electric output (~400 MWe) can be achieved with a high repetition (30 Hz) laser. This dry wall reactor concept not only reduces several difficulties associated with a liquid wall system but also enables a simple cask maintenance method for the replacement of the blanket system, which can shorten the maintenance period. The basic idea of the maintenance method for the final optics system has also been proposed. Some critical R&D issues required for this design are also discussed.

Laser-based sensor for a coolant leak detection in a nuclear reactor

NASA Astrophysics Data System (ADS)

Kim, T.-S.; Park, H.; Ko, K.; Lim, G.; Cha, Y.-H.; Han, J.; Jeong, D.-Y.

2010-08-01

Currently, the nuclear industry needs strongly a reliable detection system to continuously monitor a coolant leak during a normal operation of reactors for the ensurance of nuclear safety. In this work, we propose a new device for the coolant leak detection based on tunable diode laser spectroscopy (TDLS) by using a compact diode laser. For the feasibility experiment, we established an experimental setup consisted of a near-IR diode laser with a wavelength of about 1392 nm, a home-made multi-pass cell and a sample injection system. The feasibility test was performed for the detection of the heavy water (D2O) leaks which can happen in a pressurized heavy water reactor (PWHR). As a result, the device based on the TDLS is shown to be operated successfully in detecting a HDO molecule, which is generated from the leaked heavy water by an isotope exchange reaction between D2O and H2O. Additionally, it is suggested that the performance of the new device, such as sensitivity and stability, can be improved by adapting a cavity enhanced absorption spectroscopy and a compact DFB diode laser. We presume that this laser-based leak detector has several advantages over the conventional techniques currently employed in the nuclear power plant, such as radiation monitoring, humidity monitoring and FT-IR spectroscopy.

High-rate wastewater treatment combining a moving bed biofilm reactor and enhanced particle separation.

PubMed

Helness, H; Melin, E; Ulgenes, Y; Järvinen, P; Rasmussen, V; Odegaard, H

2005-01-01

Many cities around the world are looking for compact wastewater treatment alternatives since space for treatment plants is becoming scarce. In this paper development of a new compact, high-rate treatment concept with results from experiments in lab-scale and pilot-scale are presented. The idea behind the treatment concept is that coagulation/floc separation may be used to separate suspended and colloidal matter (resulting in > 70% organic matter removal in normal wastewater) while a high-rate biofilm process (based on Moving Bed biofilm reactors) may be used for removing low molecular weight, easily biodegradable, soluble organic matter. By using flotation for floc/biomass separation, the total residence time for a plant according to this concept will normally be < 1 hour. A cationic polymer combined with iron is used as coagulant at low dosages (i.e. 1-2 mg polymer/l, 5-10 mg Fe/l) resulting in low sludge production (compared to conventional chemical treatment) and sufficient P-removal.

Method of making a catalytic reactor for automobile

DOE Office of Scientific and Technical Information (OSTI.GOV)

Vroman, W.R.

1976-09-07

A catalytic reactor is made by providing a generally cylindrical catalytic substrate of oval transverse section and clamping the same between paired housing shells to form a housing of oval section spaced from the substrate by means of a pair of wire mesh ropes seated within a corresponding pair of grooves extending around the periphery of the substrate at axially spaced locations. Each rope is compacted from a matrix of multiple layers of resilient stainless steel knitted wire and is interlocked with the housing by means of a pair of inwardly opening channels of the housing spaced axially by anmore » inwardly projecting rib of the housing. The grooves in the substrate are pressed radially into the latter while the same is in a plastic uncured condition, thereby to compact and reinforce the grooves to withstand the localized compressional force of the ropes seated therein after the substrate is cured and hardened and clamped between the housing shells.« less

Combustion of Na 2B 4O 7 + Mg + C to synthesis B 4C powders

NASA Astrophysics Data System (ADS)

Guojian, Jiang; Jiayue, Xu; Hanrui, Zhuang; Wenlan, Li

2009-09-01

Boron carbide powder was fabricated by combustion synthesis (CS) method directly from mixed powders of borax (Na 2B 4O 7), magnesium (Mg) and carbon. The adiabatic temperature of the combustion reaction of Na 2B 4O 7 + 6 Mg + C was calculated. The control of the reactions was achieved by selecting reactant composition, relative density of powder compact and gas pressure in CS reactor. The effects of these different influential factors on the composition and morphologies of combustion products were investigated. The results show that, it is advantageous for more Mg/Na 2B 4O 7 than stoichiometric ratio in Na 2B 4O 7 + Mg + C system and high atmosphere pressure in the CS reactor to increase the conversion degree of reactants to end product. The final product with the minimal impurities' content could be fabricated at appropriate relative density of powder compact. At last, boron carbide without impurities could be obtained after the acid enrichment and distilled water washing.

Blanket activation and afterheat for the Compact Reversed-Field Pinch Reactor

NASA Astrophysics Data System (ADS)

Davidson, J. W.; Battat, M. E.

A detailed assessment has been made of the activation and afterheat for a Compact Reversed-Field Pinch Reactor (CRFPR) blanket using a two-dimensional model that included the limiter, the vacuum ducts, and the manifolds and headers for cooling the limiter and the first and second walls. Region-averaged, multigroup fluxes and prompt gamma-ray/neutron heating rates were calculated using the two-dimensional, discrete-ordinates code TRISM. Activation and depletion calculations were performed with the code FORIG using one-group cross sections generated with the TRISM region-averaged fluxes. Afterheat calculations were performed for regions near the plasma, i.e., the limiter, first wall, etc. assuming a 10-day irradiation. Decay heats were computed for decay periods up to 100 minutes. For the activation calculations, the irradiation period was taken to be one year and blanket activity inventories were computed for decay times to 4 x 10 years. These activities were also calculated as the toxicity-weighted biological hazard potential (BHP).

Advanced particle-in-cell simulation techniques for modeling the Lockheed Martin Compact Fusion Reactor

NASA Astrophysics Data System (ADS)

Welch, Dale; Font, Gabriel; Mitchell, Robert; Rose, David

2017-10-01

We report on particle-in-cell developments of the study of the Compact Fusion Reactor. Millisecond, two and three-dimensional simulations (cubic meter volume) of confinement and neutral beam heating of the magnetic confinement device requires accurate representation of the complex orbits, near perfect energy conservation, and significant computational power. In order to determine initial plasma fill and neutral beam heating, these simulations include ionization, elastic and charge exchange hydrogen reactions. To this end, we are pursuing fast electromagnetic kinetic modeling algorithms including a two implicit techniques and a hybrid quasi-neutral algorithm with kinetic ions. The kinetic modeling includes use of the Poisson-corrected direct implicit, magnetic implicit, as well as second-order cloud-in-cell techniques. The hybrid algorithm, ignoring electron inertial effects, is two orders of magnitude faster than kinetic but not as accurate with respect to confinement. The advantages and disadvantages of these techniques will be presented. Funded by Lockheed Martin.

An overview of optical diagnostics developed for the Lockheed Martin compact fusion reactor

NASA Astrophysics Data System (ADS)

Sommers, Bradley; Raymond, Anthony; Gucker, Sarah; Lockheed Martin Compact Fusion Reactor Team

2017-10-01

The T4B experiment is a linear, encapsulated ring cusp confinement device, designed to develop a physics and technology basis for a follow-on high beta machine as part of the compact fusion reactor program. Toward this end, a collection of non-invasive optical diagnostics have been developed to investigate confinement, neutral beam heating, and source behavior on the T4B device. These diagnostics include: (1) a multipoint Thomson scattering system employing a 532 nm Nd:YAG laser and high throughput spectrometer to measure 1D profiles of electron density and temperature, (2) a dispersion interferometer utilizing a continuous-wave CO2 laser (10.6 μm) to measure time resolved, line-integrated electron density, and (3) a bolometer suite utilizing four AXUV photodiodes with 64 lines of sight to generate 2D reconstructions of total radiative power and soft x-ray emission (via beryllium filters). An overview of design methods, including laser systems, detection schemes, and data analysis techniques is presented as well as results to date.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harvey, R. W.

This DOE grant supported fusion energy research, a potential long-term solution to the world's energy needs. Magnetic fusion, exemplified by confinement of very hot ionized gases, i.e., plasmas, in donut-shaped tokamak vessels is a leading approach for this energy source. Thus far, a mixture of hydrogen isotopes has produced 10's of megawatts of fusion power for seconds in a tokamak reactor at Princeton Plasma Physics Laboratory in New Jersey. The research grant under consideration, ER54684, uses computer models to aid in understanding and projecting efficacy of heating and current drive sources in the National Spherical Torus Experiment, a tokamak variant,more » at PPPL. The NSTX experiment explores the physics of very tight aspect ratio, almost spherical tokamaks, aiming at producing steady-state fusion plasmas. The current drive is an integral part of the steady-state concept, maintaining the magnetic geometry in the steady-state tokamak. CompX further developed and applied models for radiofrequency (rf) heating and current drive for applications to NSTX. These models build on a 30 year development of rf ray tracing (the all-frequencies GENRAY code) and higher dimensional Fokker-Planck rf-collisional modeling (the 3D collisional-quasilinear CQL3D code) at CompX. Two mainline current-drive rf modes are proposed for injection into NSTX: (1) electron Bernstein wave (EBW), and (2) high harmonic fast wave (HHFW) modes. Both these current drive systems provide a means for the rf to access the especially high density plasma--termed high beta plasma--compared to the strength of the required magnetic fields. The CompX studies entailed detailed modeling of the EBW to calculate the efficiency of the current drive system, and to determine its range of flexibility for driving current at spatial locations in the plasma cross-section. The ray tracing showed penetration into NSTX bulk plasma, relatively efficient current drive, but a limited ability to produce current over the whole radial plasma cross-section. The actual EBW experiment will cost several million dollars, and remains in the proposal stage. The HHFW current drive system has been experimentally implemented on NSTX, and successfully drives substantial current. The understanding of the experiment is to be accomplished in terms of general concepts of rf current drive, and also detailed modeling of the experiment which can discern the various competing processes which necessarily occur simultaneously in the experiment. An early discovery of the CompX codes, GENRAY and CQL3D, was that there could be significant interference between the neutral beam injection fast ions in the machine (injected for plasma heating) and the HHFW energy. Under many NSTX experimental conditions, power which could go to the fast ions would then be unavailable for current drive by the desired HHFW interaction with electrons. This result has been born out by experiments; the modeling helps in understanding difficulties with HHFW current drive, and has enabled adjustment of the experiment to avoid interaction with neutral beam injected fast ions thereby achieving stronger HHFW current drive. The detailed physics modeling of the various competing processes is almost always required in fusion energy plasma physics, to ensure a reasonably accurate and certain interpretation of the experiment, enabling the confident design of future, more advanced experiments and ultimately a commercial fusion reactor. More recent work entails detailed investigation of the interaction of the HHFW radiation for fast ions, accounting for the particularly large radius orbits in NSTX, and correlations between multiple HHFW-ion interactions. The spherical aspect of the NSTX experiment emphasized particular physics such as the large orbits which are present to some degree in all tokamaks, but gives clearer clues on the resulting physics phenomena since competing physics effects are reduced.« less

Experimental Measurement and Numerical Modeling of the Effective Thermal Conductivity of TRISO Fuel Compacts

DOE Office of Scientific and Technical Information (OSTI.GOV)

Folsom, Charles; Xing, Changhu; Jensen, Colby

2015-03-01

Accurate modeling capability of thermal conductivity of tristructural-isotropic (TRISO) fuel compacts is important to fuel performance modeling and safety of Generation IV reactors. To date, the effective thermal conductivity (ETC) of tristructural-isotropic (TRISO) fuel compacts has not been measured directly. The composite fuel is a complicated structure comprised of layered particles in a graphite matrix. In this work, finite element modeling is used to validate an analytic ETC model for application to the composite fuel material for particle-volume fractions up to 40%. The effect of each individual layer of a TRISO particle is analyzed showing that the overall ETC ofmore » the compact is most sensitive to the outer layer constituent. In conjunction with the modeling results, the thermal conductivity of matrix-graphite compacts and the ETC of surrogate TRISO fuel compacts have been successfully measured using a previously developed measurement system. The ETC of the surrogate fuel compacts varies between 50 and 30 W m -1 K -1 over a temperature range of 50-600°C. As a result of the numerical modeling and experimental measurements of the fuel compacts, a new model and approach for analyzing the effect of compact constituent materials on ETC is proposed that can estimate the fuel compact ETC with approximately 15-20% more accuracy than the old method. Using the ETC model with measured thermal conductivity of the graphite matrix-only material indicate that, in the composite form, the matrix material has a much greater thermal conductivity, which is attributed to the high anisotropy of graphite thermal conductivity. Therefore, simpler measurements of individual TRISO compact constituents combined with an analytic ETC model, will not provide accurate predictions of overall ETC of the compacts emphasizing the need for measurements of composite, surrogate compacts.« less

The Physics of Local Helicity Injection Non-Solenoidal Tokamak Startup

NASA Astrophysics Data System (ADS)

Redd, A. J.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Jardin, S.

2013-10-01

Non-solenoidal startup via Local Helicity Injection (LHI) uses compact current injectors to produce toroidal plasma current Ip up to 170 kA in the PEGASUS Toroidal Experiment, driven by 4-8 kA injector current on timescales of 5-20 milliseconds. Increasing the Ip buildup duration enables experimental demonstration of plasma position control on timescales relevant for high-current startup. LHI-driven discharges exhibit bursty MHD activity, apparently line-tied kinking of LHI-driven field lines, with the bursts correlating with rapid equilibrium changes, sharp Ip rises, and sharp drops in the injector impedance. Preliminary NIMROD results suggest that helical LHI-driven current channels remain coherent, with Ip increases due to reconnection between adjacent helical turns forming axisymmetric plasmoids, and corresponding sharp drops in the bias circuit impedance. The DC injector impedance is consistent with a space charge limit at low bias current and a magnetic limit at high bias current. Internal measurements show the current density profile starts strongly hollow and rapidly fills in during Ip buildup. Simulations of LHI discharges using the Tokamak Simulation Code (TSC) will provide insight into the detailed current drive mechanism and guide experiments on PEFASUS and NSTX-U. Work supported by US DOE Grants DE-FG02-96ER54375 and DE-SC0006928.

Improvement of Current Drive Efficiency in Projected FNSF Discharges

NASA Astrophysics Data System (ADS)

Prater, R.; Chan, V.; Garofalo, A.

2012-10-01

The Fusion Nuclear Science Facility - Advanced Tokamak (FNSF-AT) is envisioned as a facility that uses the tokamak approach to address the development of the AT path to fusion and fusion's energy objectives. It uses copper coils for a compact device with high βN and moderate power gain. The major radius is 2.7 m and central toroidal field is 5.44 T. Achieving the required confinement and stability at βN˜3.7 requires a current profile with negative central shear and qmin>1. Off-axis Electron Cyclotron Current Drive (ECCD), in addition to high bootstrap current fraction, can help support this current profile. Using the applied EC frequency and launch location as free parameters, a systematic study has been carried out to optimize the ECCD in the range ρ= 0.5-0.7. Using a top launch, making use of a large toroidal component to the launch direction, adjusting the vertical launch angle so that the rays propagate nearly parallel to the resonance, and adjusting the frequency for optimum total current give a high dimensionless efficiency of 0.44 for a broad ECCD profile peaked at ρ=0.7, and the driven current is 17 kA/MW for n20= 2.1 and Te= 10.3 keV locally.

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

NASA Astrophysics Data System (ADS)

Wang, Y.; Tobias, B.; Chang, Y.-T.; Yu, J.-H.; Li, M.; Hu, F.; Chen, M.; Mamidanna, M.; Phan, T.; Pham, A.-V.; Gu, J.; Liu, X.; Zhu, Y.; Domier, C. W.; Shi, L.; Valeo, E.; Kramer, G. J.; Kuwahara, D.; Nagayama, Y.; Mase, A.; Luhmann, N. C., Jr.

2017-07-01

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. Microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These have the potential to greatly advance microwave fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfvén eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today’s most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.

Millimeter-wave imaging of magnetic fusion plasmas: technology innovations advancing physics understanding

DOE PAGES

Wang, Y.; Tobias, B.; Chang, Y. -T.; ...

2017-03-14

Electron cyclotron emission (ECE) imaging is a passive radiometric technique that measures electron temperature fluctuations; and microwave imaging reflectometry (MIR) is an active radar imaging technique that measures electron density fluctuations. The microwave imaging diagnostic instruments employing these techniques have made important contributions to fusion science and have been adopted at major fusion facilities worldwide including DIII-D, EAST, ASDEX Upgrade, HL-2A, KSTAR, LHD, and J-TEXT. In this paper, we describe the development status of three major technological advancements: custom mm-wave integrated circuits (ICs), digital beamforming (DBF), and synthetic diagnostic modeling (SDM). These also have the potential to greatly advance microwavemore » fusion plasma imaging, enabling compact and low-noise transceiver systems with real-time, fast tracking ability to address critical fusion physics issues, including ELM suppression and disruptions in the ITER baseline scenario, naturally ELM-free states such as QH-mode, and energetic particle confinement (i.e. Alfven eigenmode stability) in high-performance regimes that include steady-state and advanced tokamak scenarios. Furthermore, these systems are fully compatible with today's most challenging non-inductive heating and current drive systems and capable of operating in harsh environments, making them the ideal approach for diagnosing long-pulse and steady-state tokamaks.« less

Aircraft and Bases Powered by Compact Nuclear Reactors: Solutions to Projecting Power in Highly Contested Environments and Fossil Fuel Dependence

DTIC Science & Technology

2015-05-01

pushed the depletion date past 2100.21 David Archibald, author of books and papers on climate science and a fellow at the Institute of World...Politics, does not predict explicitly the date of complete exhaustion, but he does note that humans have consumed about half of the world’s supply.22...deuterium, and lithium are plentiful on the earth and in the solar system. As far as fuel for existing and future fission reactors, uranium and

Materials interactions between the thermoelectric converter and the 5kwe reactor system

NASA Technical Reports Server (NTRS)

Ferry, P. B.

1973-01-01

The integration of a compact thermoelectric converter with a 5-kwe reactor system is described. Material interaction uncertainties study is also presented. This includes degradation of the required austenitic - refractory metal transition joint during operation at high temperatures; loss of corrosion resistance; embrittlement by the presence of hydrogen; and loss of design margin by transport of interstitial elements. Analysis and limited experimental evidence indicate that these potential materials interactions can be adequately controlled. Group 5-2 refractory metals can be utilized without unacceptable adverse effect on system reliability.

Electronics for the STEREO experiment

NASA Astrophysics Data System (ADS)

HÉLAINE, Victor; STEREO Collaboration

2017-09-01

The STEREO experiment, aiming to probe short baseline neutrino oscillations by precisely measuring reactor anti-neutrino spectrum, is currently under installation. It is located at short distance from the compact research reactor core of the Institut Laue-Langevin, Grenoble, France. Dedicated electronics, hosted in a single µTCA crate, were designed for this experiment. In this article, the electronics requirements, architecture and the performances achieved are described. It is shown how intrinsic Pulse Shape Discrimination properties of the liquid scintillator are preserved and how custom adaptable logic is used to improve the muon veto efficiency.

Physics objectives of PI3 spherical tokamak program

NASA Astrophysics Data System (ADS)

Howard, Stephen; Laberge, Michel; Reynolds, Meritt; O'Shea, Peter; Ivanov, Russ; Young, William; Carle, Patrick; Froese, Aaron; Epp, Kelly

2017-10-01

Achieving net energy gain with a Magnetized Target Fusion (MTF) system requires the initial plasma state to satisfy a set of performance goals, such as particle inventory (1021 ions), sufficient magnetic flux (0.3 Wb) to confine the plasma without MHD instability, and initial energy confinement time several times longer than the compression time. General Fusion (GF) is now constructing Plasma Injector 3 (PI3) to explore the physics of reactor-scale plasmas. Energy considerations lead us to design around an initial state of Rvessel = 1 m. PI3 will use fast coaxial helicity injection via a Marshall gun to create a spherical tokamak plasma, with no additional heating. MTF requires solenoid-free startup with no vertical field coils, and will rely on flux conservation by a metal wall. PI3 is 5x larger than SPECTOR so is expected to yield magnetic lifetime increase of 25x, while peak temperature of PI3 is expected to be similar (400-500 eV) Physics investigations will study MHD activity and the resistive and convective evolution of current, temperature and density profiles. We seek to understand the confinement physics, radiative loss, thermal and particle transport, recycling and edge physics of PI3.

Multichannel reconfigurable measurement system for hot plasma diagnostics based on GEM-2D detector

NASA Astrophysics Data System (ADS)

Wojenski, A. J.; Kasprowicz, G.; Pozniak, K. T.; Byszuk, A.; Chernyshova, M.; Czarski, T.; Jablonski, S.; Juszczyk, B.; Zienkiewicz, P.

2015-12-01

In the future magnetically confined fusion research reactors (e.g. ITER tokamak), precise determination of the level of the soft X-ray radiation of plasma with temperature above 30 keV (around 350 mln K) will be very important in plasma parameters optimization. This paper presents the first version of a designed spectrography measurement system. The system is already installed at JET tokamak. Based on the experience gained from the project, the new generation of hardware for spectrography measurements, was designed and also described in the paper. The GEM detector readout structure was changed to 2D in order to perform measurements of i.e. laser generated plasma. The hardware structure of the system was redesigned in order to provide large number of high speed input channels. Finally, this paper also covers the issue of new control software, necessary to set-up a complete system of certain complexity and perform data acquisition. The main goal of the project was to develop a new version of the system, which includes upgraded structure and data transmission infrastructure (i.e. handling large number of measurement channels, high sampling rate).

Tritiated Water on Molecular Sieve: Water Dynamics and Pressure Observations

DOE Office of Scientific and Technical Information (OSTI.GOV)

Walters, R.T.

1999-04-23

The production of fusion energy in a Tokamak using deuterium and tritium requires the safe handling and processing of exhaust gases that contain various amounts of tritium. Initial operation of the Tokamak Fusion Test Reactor (TFTR), Princeton Plasma Physics Laboratory, oxidized exhaust gases for tritium recovery or long-term storage. One of the most efficient and safest ways to contain tritiated water is to sorb it onto a pelletized 4A molecular sieve. A Disposable Molecular Sieve Bed (DMSB) was designed as a pressure vessel because of the possibility of pressure generation from the radiolysis of tritiated water on molecular sieve. Hydrogenmore » production contributes to the complexity of the containers used to transport and store tritiated water, and increases the fabrication costs. Two months after removing a DMSB from the process at TFTR, a pressure in excess of that predicted from self-radiolysis was observed. Interestingly, pressure measurements at longer times (up to 2.5 years) showed less pressure than expected. Pressure was not being generated in the DMSBs at the predicted rate. This was unexpected and prompted an investigation into the mechanism responsible for the anomalous pressure measurements.« less

Compatibility of internal transport barrier with steady-state operation in the high bootstrap fraction regime on DIII-D

DOE PAGES

Garofalo, Andrea M.; Gong, Xianzu; Grierson, Brian A.; ...

2015-11-16

Recent EAST/DIII-D joint experiments on the high poloidal beta tokamak regime in DIII-D have demonstrated fully noninductive operation with an internal transport barrier (ITB) at large minor radius, at normalized fusion performance increased by ≥30% relative to earlier work. The advancement was enabled by improved understanding of the “relaxation oscillations”, previously attributed to repetitive ITB collapses, and of the fast ion behavior in this regime. It was found that the “relaxation oscillations” are coupled core-edge modes 2 amenable to wall-stabilization, and that fast ion losses which previously dictated a large plasma-wall separation to avoid wall over-heating, can be reduced tomore » classical levels with sufficient plasma density. By using optimized waveforms of the plasma-wall separation and plasma density, fully noninductive plasmas have been sustained for long durations with excellent energy confinement quality, bootstrap fraction ≥ 80%, β N ≤ 4 , β P ≥ 3 , and β T ≥ 2%. Finally, these results bolster the applicability of the high poloidal beta tokamak regime toward the realization of a steady-state fusion reactor.« less

Analysis of a tungsten sputtering experiment in DIII-D and code/data validation of high redeposition/reduced erosion

DOE Office of Scientific and Technical Information (OSTI.GOV)

Wampler, William R.; Brooks, J. N.; Elder, J. D.

2015-03-29

We analyze a DIII-D tokamak experiment where two tungsten spots on the removable DiMES divertor probe were exposed to 12 s of attached plasma conditions, with moderate strike point temperature and density (~20 eV, ~4.5 × 10 19 m –3), and 3% carbon impurity content. Both very small (1 mm diameter) and small (1 cm diameter) deposited samples were used for assessing gross and net tungsten sputtering erosion. The analysis uses a 3-D erosion/redeposition code package (REDEP/WBC), with input from a diagnostic-calibrated near-surface plasma code (OEDGE), and with focus on charge state resolved impinging carbon ion flux and energy. Themore » tungsten surfaces are primarily sputtered by the carbon, in charge states +1 to +4. We predict high redeposition (~75%) of sputtered tungsten on the 1 cm spot—with consequent reduced net erosion—and this agrees well with post-exposure DiMES probe RBS analysis data. As a result, this study and recent related work is encouraging for erosion lifetime and non-contamination performance of tokamak reactor high-Z plasma facing components.« less

Present Status of the KSTAR Superconducting Magnet System Development

NASA Astrophysics Data System (ADS)

Kim, Keeman; H, K. Park; K, R. Park; B, S. Lim; S, I. Lee; M, K. Kim; Y, Chu; W, H. Chung; S, H. Baek; J Y, Choi; H, Yonekawa; A, Chertovskikh; Y, B. Chang; J, S. Kim; C, S. Kim; D, J. Kim; N, H. Song; K, P. Kim; Y, J. Song; I, S. Woo; W, S. Han; D, K. Lee; Y, K. Oh; K, W. Cho; J, S. Park; G, S. Lee; H, J. Lee; T, K. Ko; S, J. Lee

2004-10-01

The mission of Korea Superconducting Tokamak Advanced Research (KSTAR) project is to develop an advanced steady-state superconducting tokamak for establishing a scientific and technological basis for an attractive fusion reactor. Because one of the KSTAR mission is to achieve a steady-state operation, the use of superconducting coils is an obvious choice for the magnet system. The KSTAR superconducting magnet system consists of 16 Toroidal Field (TF) coils and 14 Poloidal Field (PF) coils. Internally-cooled Cable-In-Conduit Conductors (CICC) are put into use in both the TF and PF coil systems. The TF coil system provides a field of 3.5 T at the plasma center and the PF coil system is able to provide a flux swing of 17 V-sec. The major achievement in KSTAR magnet-system development includes the development of CICC, the development of a full-size TF model coil, the development of a coil system for background magnetic-field generation, the construction of a large-scale superconducting magnet and CICC test facility. TF and PF coils are in the stage of fabrication to pave the way for the scheduled completion of KSTAR by the end of 2006.

Fusion programs in applied plasma physics

NASA Astrophysics Data System (ADS)

1992-07-01

The Applied Plasma Physics (APP) program at General Atomics (GA) described here includes four major elements: (1) Applied Plasma Physics Theory Program, (2) Alpha Particle Diagnostic, (3) Edge and Current Density Diagnostic, and (4) Fusion User Service Center (USC). The objective of the APP theoretical plasma physics research at GA is to support the DIII-D and other tokamak experiments and to significantly advance our ability to design a commercially-attractive fusion reactor. We categorize our efforts in three areas: magnetohydrodynamic (MHD) equilibria and stability; plasma transport with emphasis on H-mode, divertor, and boundary physics; and radio frequency (RF). The objective of the APP alpha particle diagnostic is to develop diagnostics of fast confined alpha particles using the interactions with the ablation cloud surrounding injected pellets and to develop diagnostic systems for reacting and ignited plasmas. The objective of the APP edge and current density diagnostic is to first develop a lithium beam diagnostic system for edge fluctuation studies on the Texas Experimental Tokamak (TEXT). The objective of the Fusion USC is to continue to provide maintenance and programming support to computer users in the GA fusion community. The detailed progress of each separate program covered in this report period is described.

New Technique of AC drive in Tokamak using Permanent Magnets

NASA Astrophysics Data System (ADS)

Matteucci, Jackson; Zolfaghari, Ali

2013-10-01

This study investigates a new technique of capturing the rotational energy of alternating permanent magnets in order to inductively drive an alternating current in tokamak devices. The use of rotational motion bypasses many of the pitfalls seen in typical inductive and non-inductive current drives. Three specific designs are presented and assessed in the following criteria: the profile of the current generated, the RMS loop voltage generated as compared to the RMS power required to maintain it, the system's feasibility from an engineering perspective. All of the analysis has been done under ideal E&M conditions using the Maxwell 3D program. Preliminary results indicate that it is possible to produce an over 99% purely toroidal current with a RMS d Φ/dt of over 150 Tm2/s, driven by 20 MW or less of rotational power. The proposed mechanism demonstrates several key advantages including an efficient mechanical drive system, the generation of pure toroidal currents, and the potential for a quasi-steady state fusion reactor. The following quantities are presented for various driving frequencies and magnet strengths: plasma current generated, loop voltage, torque and power required. This project has been supported by DOE Funding under the SULI program.

Thermal analysis of the cryostat feed through for the ITER Tokamak TF feeder

NASA Astrophysics Data System (ADS)

Zhang, Shanwen; Song, Yuntao; Lu, Kun; Wang, Zhongwei; Zhang, Jianfeng; Qin, Yongfa

2017-04-01

In Tokamaks, the toroidal field (TF) coil feeder is an important component that is used to supply the cryogens and electrical power for the TF coils. As a part of the TF feeder, the cryostat-feed through (CFT) is subject to low temperatures of 9 and 80 K inside and room temperature of 300 K outside. Based on the features of the International Thermonuclear Experimental Reactor TF feeder, the thermal performance of the CFT under the nominal conditions is studied. Taking into account the conductive, convective and radiation heat transfer, the finite element model of the CFT is built. Transient thermal analysis is performed to determine the temperatures of the CFT on the 9th day of cooldown. The model is assessed by comparing the cooling curves of the CFT after 9 days. If the simulation and experimental results are the same, the finite element model can be considered as calibrated. The model predicts that the cooling time will be approximately 26 days and the temperature distribution and heat load of the main components are obtained when the CFT reaches thermal equilibrium. This study provides a valid quantitative characterization of the CFT design.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kaita, Robert; Boyle, Dennis; Gray, Timothy

Liquid metal walls have been proposed to address the first wall challenge for fusion reactors. The Lithium Tokamak Experiment (LTX) at the Princeton Plasma Physics Laboratory (PPPL) is the first magnetic confinement device to have liquid metal plasma-facing components (PFC's) that encloses virtually the entire plasma. In the Current Drive Experiment-Upgrade (CDX-U), a predecessor to LTX at PPPL, the highest improvement in energy confinement ever observed in Ohmically-heated tokamak plasmas was achieved with a toroidal liquid lithium limiter. The LTX extends this liquid lithium PFC by using a conducting conformal shell that almost completely surrounds the plasma. By heating themore » shell, a lithium coating on the plasma-facing side can be kept liquefied. A consequence of the low-recycling conditions from liquid lithium walls is the need for efficient plasma fueling. For this purpose, a molecular cluster injector is being developed. Future plans include the installation of a neutral beam for core plasma fueling, and also ion temperature measurements using charge-exchange recombination spectroscopy. Low edge recycling is also predicted to reduce temperature gradients that drive drift wave turbulence. Gyrokinetic simulations are in progress to calculate fluctuation levels and transport for LTX plasmas, and new fluctuation diagnostics are under development to test these predictions. __________________________________________________« less

Automated Characterization of Rotating MHD Modes and Subsequent Locking in a Tokamak

NASA Astrophysics Data System (ADS)

Riquezes, Juan; Sabbagh, Steven; Berkery, Jack

2016-10-01

Disruption avoidance in tokamaks is highly desired to maintain steady plasma operation, and is critical for future reactor-scale devices, such as ITER, to avoid potential damage to device components. This high priority research is being conducted at PPPL by analyzing data from NSTX and its upgrade, NSTX-U. A key cause of disruptions is the physical event chain that comprises the appearance of rotating MHD modes, their slowing by resonant field drag mechanisms, and their subsequent locking. The present research aims to define algorithms to automatically find and characterize such physical event chains in the machine database. Characteristics such as identification of a mode locking time based on a loss of torque balance and bifurcation of the mode rotation frequency are examined to determine the reliability of such events in predicting disruptions. A goal is to detect such behavior as early as possible during a plasma discharge, and to further examine potential ways to forecast it. This capability could be used to provide a warning to use active mode control as a disruption avoidance mechanism, or to trigger a controlled plasma shutdown if desired. Supported by US DOE Contracts DE-FG02-99ER54524 and DE-AC02-09CH11466.

Comparison of silver release predictions using PARFUME with results from the AGR-2 irradiation experiment

DOE Office of Scientific and Technical Information (OSTI.GOV)

Collin, Blaise P.; Demkowicz, Paul A.; Baldwin, Charles A.

2016-11-01

The PARFUME (PARticle FUel ModEl) code was used to predict silver release from tristructural isotropic (TRISO) coated fuel particles and compacts during the second irradiation experiment (AGR-2) of the Advanced Gas Reactor Fuel Development and Qualification program. The PARFUME model for the AGR-2 experiment used the fuel compact volume average temperature for each of the 559 days of irradiation to calculate the release of fission product silver from a representative particle for a select number of AGR-2 compacts and individual fuel particles containing either mixed uranium carbide/oxide (UCO) or 100% uranium dioxide (UO2) kernels. Post-irradiation examination (PIE) measurements were performedmore » to provide data on release of silver from these compacts and individual fuel particles. The available experimental fractional releases of silver were compared to their corresponding PARFUME predictions. Preliminary comparisons show that PARFUME under-predicts the PIE results in UCO compacts and is in reasonable agreement with experimental data for UO2 compacts. The accuracy of PARFUME predictions is impacted by the code limitations in the modeling of the temporal and spatial distributions of the temperature across the compacts. Nevertheless, the comparisons on silver release lie within the same order of magnitude.« less

Demonstration Tokamak Hybrid Reactor (DTHR) blanket design study, December 1978

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1978-01-01

This work represents only the second iteration of the conceptual design of a DTHR blanket; consequently, a number of issues important to a detailed blanket design have not yet been evaluated. The most critical issues identified are those of two-phase flow maldistribution, flow instabilities, flow stratification for horizontal radial inflow of boiling water, fuel rod vibrations, corrosion of clad and structural materials by high quality steam, fretting and cyclic loads. Approaches to minimizing these problems are discussed and experimental testing with flow mock-ups is recommended. These implications on a commercial blanket design are discussed and critical data needs are identified.

Conceptual design of the cryogenic system and estimation of the recirculated power for CFETR

NASA Astrophysics Data System (ADS)

Liu, Xiaogang; Qiu, Lilong; Li, Junjun; Wang, Zhaoliang; Ren, Yong; Wang, Xianwei; Li, Guoqiang; Gao, Xiang; Bi, Yanfang

2017-01-01

The China Fusion Engineering Test Reactor (CFETR) is the next tokamak in China’s roadmap for realizing commercial fusion energy. The CFETR cryogenic system is crucial to creating and maintaining operational conditions for its superconducting magnet system and thermal shields. The preliminary conceptual design of the CFETR cryogenic system has been carried out with reference to that of ITER. It will provide an average capacity of 75 to 80 kW at 4.5 K and a peak capacity of 1300 kW at 80 K. The electric power consumption of the cryogenic system is estimated to be 24 MW, and the gross building area is about 7000 m2. The relationships among the auxiliary power consumed by the cryogenic system, the fusion power gain and the recirculated power of CFETR are discussed, with the suggestion that about 52% of the electric power produced by CFETR in phase II must be recirculated to run the fusion test reactor.

Radioactivity measurements of ITER materials using the TFTR D-T neutron field

NASA Astrophysics Data System (ADS)

Kumar, A.; Abdou, M. A.; Barnes, C. W.; Kugel, H. W.

1994-06-01

The availability of high D-T fusion neutron yields at TFTR has provided a useful opportunity to directly measure D-T neutron-induced radioactivity in a realistic tokamak fusion reactor environment for materials of vital interest to ITER. These measurements are valuable for characterizing radioactivity in various ITER candidate materials, for validating complex neutron transport calculations, and for meeting fusion reactor licensing requirements. The radioactivity measurements at TFTR involve potential ITER materials including stainless steel 316, vanadium, titanium, chromium, silicon, iron, cobalt, nickel, molybdenum, aluminum, copper, zinc, zirconium, niobium, and tungsten. Small samples of these materials were irradiated close to the plasma and just outside the vacuum vessel wall of TFTR, locations of different neutron energy spectra. Saturation activities for both threshold and capture reactions were measured. Data from dosimetric reactions have been used to obtain preliminary neutron energy spectra. Spectra from the first wall were compared to calculations from ITER and to measurements from accelerator-based tests.

ADVANTG An Automated Variance Reduction Parameter Generator, Rev. 1

DOE Office of Scientific and Technical Information (OSTI.GOV)

Mosher, Scott W.; Johnson, Seth R.; Bevill, Aaron M.

2015-08-01

The primary objective of ADVANTG is to reduce both the user effort and the computational time required to obtain accurate and precise tally estimates across a broad range of challenging transport applications. ADVANTG has been applied to simulations of real-world radiation shielding, detection, and neutron activation problems. Examples of shielding applications include material damage and dose rate analyses of the Oak Ridge National Laboratory (ORNL) Spallation Neutron Source and High Flux Isotope Reactor (Risner and Blakeman 2013) and the ITER Tokamak (Ibrahim et al. 2011). ADVANTG has been applied to a suite of radiation detection, safeguards, and special nuclear materialmore » movement detection test problems (Shaver et al. 2011). ADVANTG has also been used in the prediction of activation rates within light water reactor facilities (Pantelias and Mosher 2013). In these projects, ADVANTG was demonstrated to significantly increase the tally figure of merit (FOM) relative to an analog MCNP simulation. The ADVANTG-generated parameters were also shown to be more effective than manually generated geometry splitting parameters.« less

Synthesis gas and olefins from the catalytic autothermal reforming of volatile and non-volatile liquids

NASA Astrophysics Data System (ADS)

Dreyer, Bradon Justin

2007-12-01

The research presented in this thesis develops an understanding of a clean energy process technology, catalytic partial oxidation (CPO). CPO is a process in which a carbon containing fuel, such as a hydrocarbon, is passed over a noble metal catalyst (e.g. rhodium and platinum) to efficiently generate synthesis gas (H2 and CO) and olefins (e.g. ethylene and propylene) in millisecond contact times. Chapter 1 introduces CPO and compares this technology with conventional methods for synthesis gas and olefin production. CPO has several advantages over the traditional synthesis gas and olefin production methods. One advantage includes autothermal operation, requiring no external heat input from furnaces or heat exchangers. Autothermal operation allows these reactors to be built compactly. The short contact-times associated with CPO further enable for high throughput in relatively small reactor systems, and more compact reactors typically translate to faster response times if transient operation is required. Nobel metal based CPO catalysts are also resistant to deactivation, resulting in less catalyst replacement, regeneration, and maintenance, and an increase in operating efficiency. An overview of the many applications of the chemicals produced from CPO is also presented in Chapter 1. The chemicals produced are crucial in generating valuable chemical intermediates that are eventually incorporated in consumer products, medical devices, building structures, and fertilizers. Additionally, H2 can be used as a source of energy in mobile fuel applications. Fuel cells convert H2 and O2 into electricity and water at higher efficiencies than thermal engine generators. Due to the difficulties in H2 storage, these more efficient energy generators are dependent on hydrogen obtained from synthesis gas production in compact, portable fuel reformers, such as CPO reactors. Furthermore, H2 and CO can be used in reducing environmentally harmful emissions. Particularly, the implementation of NOx traps and hydrogen into diesel engines has shown potential in reducing NOx emissions into the environment. Both concepts are dependent on synthesis gas generated from portable, compact fuel reformers, such as CPO reactors. Chapter 1 also reviews previous research in CPO, along with several important experimental parameters, and outlines the remaining research directions in the remaining chapters. In Chapter 2, steam addition to the CPO of higher hydrocarbons was explored over rhodium-coated ceramic foam supports at millisecond contact times. Steam addition to the CPO of n-decane and n-hexadecane in air produced considerably higher H2 and CO2 and lower olefin and CO selectivities than traditional CPO. For steam to carbon feed ratios from 0.0 to 4.0, the reactor operated autothermally, and the H2 to CO product ratio increased from ˜1.0 to ˜4.0, which is essentially the equilibrium product composition near synthesis gas stoichiometry (C/O ˜1) at contact times of ˜7 milliseconds. In fuel-rich feeds exceeding the synthesis gas ratio (C/O > 1), steam addition suppressed olefins, promoted synthesis gas and water-gas shift products, and reduced catalyst surface carbon. Furthermore, steam addition to the CPO of the military fuel JP-8 was performed successfully, also increasing H2 and suppressing olefins. (Abstract shortened by UMI.)

Nuclear design of a vapor core reactor for space nuclear propulsion

NASA Astrophysics Data System (ADS)

Dugan, Edward T.; Watanabe, Yoichi; Kuras, Stephen A.; Maya, Isaac; Diaz, Nils J.

1993-01-01

Neutronic analysis methodology and results are presented for the nuclear design of a vapor core reactor for space nuclear propulsion. The Nuclear Vapor Thermal Reactor (NVTR) Rocket Engine uses modified NERVA geometry and systems which the solid fuel replaced by uranium tetrafluoride vapor. The NVTR is an intermediate term gas core thermal rocket engine with specific impulse in the range of 1000-1200 seconds; a thrust of 75,000 lbs for a hydrogen flow rate of 30 kg/s; average core exit temperatures of 3100 K to 3400 K; and reactor thermal powers of 1400 to 1800 MW. Initial calculations were performed on epithermal NVTRs using ZrC fuel elements. Studies are now directed at thermal NVTRs that use fuel elements made of C-C composite. The large ZrC-moderated reactors resulted in thrust-to-weight ratios of only 1 to 2; the compact C-C composite systems yield thrust-to-weight ratios of 3 to 5.

Determination of Trace Concentration in TMD Detectors using PGAA

NASA Astrophysics Data System (ADS)

Tomandl, I.; Viererbl, L.; Kudějová, P.; Lahodová, Z.; Klupák, V.; Fikrle, M.

2015-05-01

Transmutation detectors could be alternative to the traditional activation detector method for neutron fluence dosimetry at power nuclear reactors. This new method require an isotopically highly-sensitive, non-destructive in sense of compactness as well as isotopic content, precise and standardly used analytical method for trace concentration determination. The capability of Prompt Gamma-ray Activation Analysis (PGAA) for determination of trace concentrations of transmuted stable nuclides in the metallic foils of Ni, Au, Cu and Nb, which were irradiated for 21 days in the reactor core at the LVR-15 research reactor in Řež, is reported. The PGAA measurements of these activation foils were performed at the PGAA facility at Forschungs-Neutronenquelle Heinz Maier-Leibnitz (FRMII) in Garching.

Mass tracking and material accounting in the integral fast reactor (IFR)

DOE Office of Scientific and Technical Information (OSTI.GOV)

Orechwa, Y.; Adams, C.H.; White, A.M.

1991-01-01

This paper reports on the Integral Fast Reactor (IFR) which is a generic advanced liquid metal cooled reactor concept being developed at Argonne National Laboratory. There are a number of technical features of the IFR which contribute to its potential as a next-generation reactor. These are associated with large safety margins with regard to off-normal events involving the heat transport system, and the use of metallic fuel which makes possible the utilization of innovative fuel cycle processes. The latter feature permits fuel cycle closure with compact, low-cost reprocessing facilities, collocated with the reactor plant. These primary features are being demonstratedmore » in the facilities at ANL-West, utilizing Experimental Breeder Reactor II and the associated Fuel Cycle Facility (FCF) as an IFR prototype. The demonstration of this IFR prototype includes the design and implementation of the Mass-tracking System (MTG). In this system, data from the operations of the FCF, including weights and batch-process parameters, are collected and maintained by the MTG running on distributed workstations.« less

Feasibility study of a magnetic fusion production reactor

NASA Astrophysics Data System (ADS)

Moir, R. W.

1986-12-01

A magnetic fusion reactor can produce 10.8 kg of tritium at a fusion power of only 400 MW —an order of magnitude lower power than that of a fission production reactor. Alternatively, the same fusion reactor can produce 995 kg of plutonium. Either a tokamak or a tandem mirror production plant can be used for this purpose; the cost is estimated at about 1.4 billion (1982 dollars) in either case. (The direct costs are estimated at 1.1 billion.) The production cost is calculated to be 22,000/g for tritium and 260/g for plutonium of quite high purity (1%240Pu). Because of the lack of demonstrated technology, such a plant could not be constructed today without significant risk. However, good progress is being made in fusion technology and, although success in magnetic fusion science and engineering is hard to predict with assurance, it seems possible that the physics basis and much of the needed technology could be demonstrated in facilities now under construction. Most of the remaining technology could be demonstrated in the early 1990s in a fusion test reactor of a few tens of megawatts. If the Magnetic Fusion Energy Program constructs a fusion test reactor of approximately 400 MW of fusion power as a next step in fusion power development, such a facility could be used later as a production reactor in a spinoff application. A construction decision in the late 1980s could result in an operating production reactor in the late 1990s. A magnetic fusion production reactor (MFPR) has four potential advantages over a fission production reactor: (1) no fissile material input is needed; (2) no fissioning exists in the tritium mode and very low fissioning exists in the plutonium mode thus avoiding the meltdown hazard; (3) the cost will probably be lower because of the smaller thermal power required; (4) and no reprocessing plant is needed in the tritium mode. The MFPR also has two disadvantages: (1) it will be more costly to operate because it consumes rather than sells electricity, and (2) there is a risk of not meeting the design goals.

PREPARATION OF HIGH-DENSITY, COMPACTIBLE THORIUM OXIDE PARTICLES

DOEpatents

McCorkle, K.H.; Kleinsteuber, A.T.; Schilling, C.E.; Dean, O.C.

1962-05-22

A method is given for preparing millimeter-size, highdensity thorium oxide particles suitable for fabrication into nuclear reactor feel elements by means of vibratory compaction. A thorium oxide gel containing 3.7 to 7 weight per cent residual volatile nitrate and water is prepared by drying a thorium oxide sol. The gel is then slowly heated to a temperature of about 450DEC, and the resulting gel fragments are calcined. The starting sol is prepared by repeated dispersion of oxalate-source thorium oxide in a nitrate system or by dispersion of steam-denitrated thorium oxide in water. (AEC)

Investigation of a Possibility of Chromium-51 Accumulation in the SM-3 Reactor to Fabricate a Neutrino Source

NASA Astrophysics Data System (ADS)

Romanov, E. G.; Gavrin, V. N.; Tarasov, V. A.; Malkov, A. P.; Kupriyanov, A. V.; Danshin, S. N.; Veretenkin, E. P.

2017-01-01

Compact high intensity neutrino sources based on 51Cr isotope are demanded for very short baseline neutrino experiments. In particular, a 3 MCi 51Cr neutrino source is needed for the experiment BEST on search for transitions of electron neutrinos to sterile states. The paper presents the results of the analysis of options of the irradiation of highly enriched 50Cr in the existing trap of thermal neutrons of high-flux reactor SM-3, as well as using the most promising variants of the trap after upcoming reconstruction of the reactor. It is shown that it is possible to to obtain the intensity of 51Cr up to 3.85 MCi at the end of irradiation of 50Cr enriched to 97% in the high-flux reactor SM-3 of the JSC “SSC NIIAR”.

Plasma nitriding monitoring reactor: A model reactor for studying plasma nitriding processes using an active screen

DOE Office of Scientific and Technical Information (OSTI.GOV)

Hamann, S., E-mail: hamann@inp-greifswald.de; Röpcke, J.; Börner, K.

2015-12-15

A laboratory scale plasma nitriding monitoring reactor (PLANIMOR) has been designed to study the basics of active screen plasma nitriding (ASPN) processes. PLANIMOR consists of a tube reactor vessel, made of borosilicate glass, enabling optical emission spectroscopy (OES) and infrared absorption spectroscopy. The linear setup of the electrode system of the reactor has the advantages to apply the diagnostic approaches on each part of the plasma process, separately. Furthermore, possible changes of the electrical field and of the heat generation, as they could appear in down-scaled cylindrical ASPN reactors, are avoided. PLANIMOR has been used for the nitriding of steelmore » samples, achieving similar results as in an industrial scale ASPN reactor. A compact spectrometer using an external cavity quantum cascade laser combined with an optical multi-pass cell has been applied for the detection of molecular reaction products. This allowed the determination of the concentrations of four stable molecular species (CH{sub 4}, C{sub 2}H{sub 2}, HCN, and NH{sub 3}). With the help of OES, the rotational temperature of the screen plasma could be determined.« less

ADAPTATION OF CRACK GROWTH DETECTION TECHNIQUES TO US MATERIAL TEST REACTORS

DOE Office of Scientific and Technical Information (OSTI.GOV)

A. Joseph Palmer; Sebastien P. Teysseyre; Kurt L. Davis

2015-04-01

A key component in evaluating the ability of Light Water Reactors to operate beyond 60 years is characterizing the degradation of materials exposed to radiation and various water chemistries. Of particular concern is the response of reactor materials to Irradiation Assisted Stress Corrosion Cracking (IASCC). Some test reactors outside the United States, such as the Halden Boiling Water Reactor (HBWR), have developed techniques to measure crack growth propagation during irradiation. The basic approach is to use a custom-designed compact loading mechanism to stress the specimen during irradiation, while the crack in the specimen is monitored in-situ using the Direct Currentmore » Potential Drop (DCPD) method. In 2012 the US Department of Energy commissioned the Idaho National Laboratory and the MIT Nuclear Reactor Laboratory (MIT NRL) to take the basic concepts developed at the HBWR and adapt them to a test rig capable of conducting in-pile IASCC tests in US Material Test Reactors. The first two and half years of the project consisted of designing and testing the loader mechanism, testing individual components of the in-pile rig and electronic support equipment, and autoclave testing of the rig design prior to insertion in the MIT Reactor. The load was applied to the specimen by means of a scissor like mechanism, actuated by a miniature metal bellows driven by pneumatic pressure and sized to fit within the small in-core irradiation volume. In addition to the loader design, technical challenges included developing robust connections to the specimen for the applied current and voltage measurements, appropriate ceramic insulating materials that can endure the LWR environment, dealing with the high electromagnetic noise environment of a reactor core at full power, and accommodating material property changes in the specimen, due primarily to fast neutron damage, which change the specimen resistance without additional crack growth. The project culminated with an in-pile demonstration at the MIT Reactor. The test rig and associated support equipment were used to apply loads to a representative Compact Tensile specimen during one MITR operating cycle, while measuring crack growth using the DCPD method. Although the test period was short (approximately 70 days), and the accumulated neutron dose relatively small, successful operation of the test rig was demonstrated. The specimen was cycled more than 8000 times (more than would be typical for a long term IASCC test), which was sufficient to propagate a crack of over 2 mm.« less

Microstructural evolution of nanochannel CrN films under ion irradiation at elevated temperature and post-irradiation annealing

NASA Astrophysics Data System (ADS)

Tang, Jun; Hong, Mengqing; Wang, Yongqiang; Qin, Wenjing; Ren, Feng; Dong, Lan; Wang, Hui; Hu, Lulu; Cai, Guangxu; Jiang, Changzhong

2018-03-01

High-performance radiation tolerance materials are crucial for the success of future advanced nuclear reactors. In this paper, we present a further investigation that the "vein-like" nanochannel films can enhance radiation tolerance under ion irradiation at high temperature and post-irradiation annealing. The chromium nitride (CrN) nanochannel films with different nanochannel densities and the compact CrN film are chosen as a model system for these studies. Microstructural evolution of these films were investigated using Transmission Electron Microscopy (TEM), Scanning Electron Microscopy (SEM), Elastic Recoil Detection (ERD) and Grazing Incidence X-ray Diffraction (GIXRD). Under the high fluence He+ ion irradiation at 500 °C, small He bubbles with low bubble densities are observed in the irradiated nanochannel CrN films, while the aligned large He bubbles, blistering and texture reconstruction are found in the irradiated compact CrN film. For the heavy Ar2+ ion irradiation at 500 °C, the microstructure of the nanochannel CrN RT film is more stable than that of the compact CrN film due to the effective releasing of defects via the nanochannel structure. Under the He+ ion irradiation and subsequent annealing, compared with the compact film, the nanochannel films have excellent performance for the suppression of He bubble growth and possess the strong microstructural stability. Basing on the analysis on the sizes and number densities of bubbles as well as the concentrations of He retained in the nanochannel CrN films and the compact CrN film under different experimental conditions, potential mechanism for the enhanced radiation tolerance are discussed. Nanochannels play a crucial role on the release of He/defects under ion irradiation. We conclude that the tailored "vein-like" nanochannel structure may be used as advanced radiation tolerance materials for future nuclear reactors.

An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

DOE Office of Scientific and Technical Information (OSTI.GOV)

Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.

AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry for TRISO fuel compacts across a burnup range of approximately 10 to 20% FIMA and also validate the approach used in the physics simulation of the AGR 1 experiment.« less

An analysis of nuclear fuel burnup in the AGR-1 TRISO fuel experiment using gamma spectrometry, mass spectrometry, and computational simulation techniques

DOE PAGES

Harp, Jason M.; Demkowicz, Paul A.; Winston, Philip L.; ...

2014-09-03

AGR 1 was the first in a series of experiments designed to test US TRISO fuel under high temperature gas-cooled reactor irradiation conditions. This experiment was irradiated in the Advanced Test Reactor (ATR) at Idaho National Laboratory (INL) and is currently undergoing post irradiation examination (PIE) at INL and Oak Ridge National Laboratory. One component of the AGR 1 PIE is the experimental evaluation of the burnup of the fuel by two separate techniques. Gamma spectrometry was used to non destructively evaluate the burnup of all 72 of the TRISO fuel compacts that comprised the AGR 1 experiment. Two methodsmore » for evaluating burnup by gamma spectrometry were developed, one based on the Cs 137 activity and the other based on the ratio of Cs 134 and Cs 137 activities. Burnup values determined from both methods compared well with the values predicted from simulations. The highest measured burnup was 20.1% FIMA for the direct method and 20.0% FIMA for the ratio method (compared to 19.56% FIMA from simulations). An advantage of the ratio method is that the burnup of the cylindrical fuel compacts can determined in small (2.5 mm) axial increments and an axial burnup profile can be produced. Destructive chemical analysis by inductively coupled mass spectrometry (ICP MS) was then performed on selected compacts that were representative of the expected range of fuel burnups in the experiment to compare with the burnup values determined by gamma spectrometry. The compacts analyzed by mass spectrometry had a burnup range of 19.3% FIMA to 10.7% FIMA. The mass spectrometry evaluation of burnup for the four compacts agreed well with the gamma spectrometry burnup evaluations and the expected burnup from simulation. For all four compacts analyzed by mass spectrometry, the maximum range in the three experimentally determined values and the predicted value was 6% or less. Furthermore, the results confirm the accuracy of the nondestructive burnup evaluation from gamma spectrometry for TRISO fuel compacts across a burnup range of approximately 10 to 20% FIMA and also validate the approach used in the physics simulation of the AGR 1 experiment.« less

Critical Design Issues of Tokamak Cooling Water System of ITER's Fusion Reactor

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kim, Seokho H; Berry, Jan

U.S. ITER is responsible for the design, engineering, and procurement of the Tokamak Cooling Water System (TCWS). The TCWS transfers heat generated in the Tokamak to cooling water during nominal pulsed operation 850 MW at up to 150 C and 4.2 MPa water pressure. This water contains radionuclides because impurities (e.g., tritium) diffuse from in-vessel components and the vacuum vessel by water baking at 200 240 C at up to 4.4MPa, and corrosion products become activated by neutron bombardment. The system is designated as safety important class (SIC) and will be fabricated to comply with the French Order concerning nuclearmore » pressure equipment (December 2005) and the EU Pressure Equipment Directive using ASME Section VIII, Div 2 design codes. The complexity of the TCWS design and fabrication presents unique challenges. Conceptual design of this one-of-a-kind cooling system has been completed with several issues that need to be resolved to move to next stage of the design. Those issues include flow balancing between over hundreds of branch pipelines in parallel to supply cooling water to blankets, determination of optimum flow velocity while minimizing the potential for cavitation damage, design for freezing protection for cooling water flowing through cryostat (freezing) environment, requirements for high-energy piping design, and electromagnetic impact to piping and components. Although the TCWS consists of standard commercial components such as piping with valves and fittings, heat exchangers, and pumps, complex requirements present interesting design challenges. This paper presents a brief description of TCWS conceptual design and critical design issues that need to be resolved.« less

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finley, V.L.; Wieczorek, M.A.

This report gives the results of the environmental activities and monitoring programs at the Princeton Plasma Physics Laboratory (PPPL) for CY94. The report is prepared to provide the US Department of Energy (DOE) and the public with information on the level of radioactive and nonradioactive pollutants, if any, added to the environment as a result of PPPL operations, as well as environmental initiatives, assessments, and programs that were undertaken in 1994. The objective of the Annual Site Environmental Report is to document evidence that PPPL`s environmental protection programs adequately protect the environment and the public health. The Princeton Plasma Physicsmore » Laboratory has engaged in fusion energy research since 195 1. The long-range goal of the US Magnetic Fusion Energy Research Program is to develop and demonstrate the practical application of fusion power as an alternate energy source. In 1994, PPPL had one of its two large tokamak devices in operation-the Tokamak Fusion Test Reactor (TFTR). The Princeton Beta Experiment-Modification or PBX-M completed its modifications and upgrades and resumed operation in November 1991 and operated periodically during 1992 and 1993; it did not operate in 1994 for funding reasons. In December 1993, TFTR began conducting the deuterium-tritium (D-T) experiments and set new records by producing over ten @on watts of energy in 1994. The engineering design phase of the Tokamak Physics Experiment (T?X), which replaced the cancelled Burning Plasma Experiment in 1992 as PPPL`s next machine, began in 1993 with the planned start up set for the year 2001. In December 1994, the Environmental Assessment (EA) for the TFTR Shutdown and Removal (S&R) and TPX was submitted to the regulatory agencies, and a finding of no significant impact (FONSI) was issued by DOE for these projects.« less

Overview of physics research on the TCV tokamak

NASA Astrophysics Data System (ADS)

Fasoli, A.; TCV Team

2009-10-01

The Tokamak à Configuration Variable (TCV) tokamak is equipped with high-power (4.5 MW), real-time-controllable EC systems and flexible shaping, and plays an important role in fusion research by broadening the parameter range of reactor relevant regimes, by investigating tokamak physics questions and by developing new control tools. Steady-state discharges are achieved, in which the current is entirely self-generated through the bootstrap mechanism, a fundamental ingredient for ITER steady-state operation. The discharge remains quiescent over several current redistribution times, demonstrating that a self-consistent, 'bootstrap-aligned' equilibrium state is possible. Electron internal transport barrier regimes sustained by EC current drive have also been explored. MHD activity is shown to be crucial in scenarios characterized by large and slow oscillations in plasma confinement, which in turn can be modified by small Ohmic current perturbations altering the barrier strength. In studies of the relation between anomalous transport and plasma shape, the observed dependences of the electron thermal diffusivity on triangularity (direct) and collisionality (inverse) are qualitatively reproduced by non-linear gyro-kinetic simulations and shown to be governed by TEM turbulence. Parallel SOL flows are studied for their importance for material migration. Flow profiles are measured using a reciprocating Mach probe by changing from lower to upper single-null diverted equilibria and shifting the plasmas vertically. The dominant, field-direction-dependent Pfirsch-Schlüter component is found to be in good agreement with theoretical predictions. A field-direction-independent component is identified and is consistent with flows generated by transient over-pressure due to ballooning-like interchange turbulence. Initial high-resolution infrared images confirm that ELMs have a filamentary structure, while fast, localized radiation measurements reveal that ELM activity first appears in the X-point region. Real time control techniques are currently being applied to EC multiple independent power supplies and beam launchers, e.g. to control the plasma current in fully non-inductive conditions, and the plasma elongation through current broadening by far-off-axis heating at constant shaping field.

Comparison of Homogeneous and Heterogeneous CFD Fuel Models for Phase I of the IAEA CRP on HTR Uncertainties Benchmark

DOE Office of Scientific and Technical Information (OSTI.GOV)

Gerhard Strydom; Su-Jong Yoon

2014-04-01

Computational Fluid Dynamics (CFD) evaluation of homogeneous and heterogeneous fuel models was performed as part of the Phase I calculations of the International Atomic Energy Agency (IAEA) Coordinate Research Program (CRP) on High Temperature Reactor (HTR) Uncertainties in Modeling (UAM). This study was focused on the nominal localized stand-alone fuel thermal response, as defined in Ex. I-3 and I-4 of the HTR UAM. The aim of the stand-alone thermal unit-cell simulation is to isolate the effect of material and boundary input uncertainties on a very simplified problem, before propagation of these uncertainties are performed in subsequent coupled neutronics/thermal fluids phasesmore » on the benchmark. In many of the previous studies for high temperature gas cooled reactors, the volume-averaged homogeneous mixture model of a single fuel compact has been applied. In the homogeneous model, the Tristructural Isotropic (TRISO) fuel particles in the fuel compact were not modeled directly and an effective thermal conductivity was employed for the thermo-physical properties of the fuel compact. On the contrary, in the heterogeneous model, the uranium carbide (UCO), inner and outer pyrolytic carbon (IPyC/OPyC) and silicon carbide (SiC) layers of the TRISO fuel particles are explicitly modeled. The fuel compact is modeled as a heterogeneous mixture of TRISO fuel kernels embedded in H-451 matrix graphite. In this study, a steady-state and transient CFD simulations were performed with both homogeneous and heterogeneous models to compare the thermal characteristics. The nominal values of the input parameters are used for this CFD analysis. In a future study, the effects of input uncertainties in the material properties and boundary parameters will be investigated and reported.« less

PREPARATION OF UO$sub 2$ FOR NUCLEAR REACTOR FUEL PELLETS

DOEpatents

Googin, J.M.

1962-06-01

A method is given for preparing high-density UO/sub 2/ compacts. An aqueous uranyl fluoride solution is contacted with an aqueous ammonium hydroxide solution at an ammonium to-uranium ratio of 25: 1 to 30:1 to form a precipitate. The precipitate is separated from the- mother liquor, dried, and contacted with steam at a uniform temperature within the range of 400 to 650 deg C to produce U/ sub 3/O/sub 8/. The U/sub 3/O/sub 8/ is red uced to UO/sub 2/ with hydrogen at a uniform temperature within the range of 550 to 600 deg C. The UO/sub 2/ is then compressed into compacts and sintered. High-density compacts are fabricated to close tolerances without use of a binder and without machining or grinding. (AEC)

HRB-22 preirradiation thermal analysis

DOE Office of Scientific and Technical Information (OSTI.GOV)

Acharya, R.; Sawa, K.

1995-05-01

This report describes the preirradiation thermal analysis of the HRB-22 capsule designed for irradiation in the removable beryllium (RB) position of the High Flux Isotope Reactor (HFIR) at Oak Ridge National Laboratory (ORNL). CACA-2 a heavy isotope and fission product concentration calculational code for experimental irradiation capsules was used to determine time dependent fission power for the fuel compacts. The Heat Engineering and Transfer in Nine Geometries (HEATING) computer code, version 7.2, was used to solve the steady-state heat conduction problem. The diameters of the graphite fuel body that contains the compacts and the primary pressure vessel were selected suchmore » that the requirements of running the compacts at an average temperature of < 1,250 C and not exceeding a maximum fuel temperature of 1,350 C was met throughout the four cycles of irradiation.« less

The Joint European Torus (JET)

NASA Astrophysics Data System (ADS)

Rebut, Paul-Henri

2017-02-01

This paper addresses the history of JET, the Tokamak that reached the highest performances and the experiment that so far came closest to the eventual goal of a fusion reactor. The reader must be warned, however, that this document is not a comprehensive study of controlled thermonuclear fusion or even of JET. The next step on this road, the ITER project, is an experimental reactor. Actually, several prototypes will be required before a commercial reactor can be built. The fusion history is far from been finalised. JET is still in operation some 32 years after the first plasma and still has to provide answers to many questions before ITER takes the lead on research. Some physical interpretations of the observed phenomena, although coherent, are still under discussion. This paper also recalls some basic physics concepts necessary to the understanding of confinement: a knowledgeable reader can ignore these background sections. This fascinating story, comprising successes and failures, is imbedded in the complexities of twentieth and the early twenty-first centuries at a time when world globalization is evolving and the future seems loaded with questions. The views here expressed on plasma confinement are solely those of the author. This is especially the case for magnetic turbulence, for which other scientists may have different views.

Extruder system for high-throughput/steady-state hydrogen ice supply and application for pellet fueling of reactor-scale fusion experiments

DOE Office of Scientific and Technical Information (OSTI.GOV)

Combs, S.K.; Foust, C.R.; Qualls, A.L.

Pellet injection systems for the next-generation fusion devices, such as the proposed International Thermonuclear Experimental Reactor (ITER), will require feed systems capable of providing a continuous supply of hydrogen ice at high throughputs. A straightforward concept in which multiple extruder units operate in tandem has been under development at the Oak Ridge National Laboratory. A prototype with three large-volume extruder units has been fabricated and tested in the laboratory. In experiments, it was found that each extruder could provide volumetric ice flow rates of up to {approximately}1.3 cm{sup 3}/s (for {approximately}10 s), which is sufficient for fueling fusion reactors atmore » the gigawatt power level. With the three extruders of the prototype operating in sequence, a steady rate of {approximately}0.33 cm{sup 3}/s was maintained for a duration of 1 h. Even steady-state rates approaching the full ITER design value ({approximately}1 cm{sup 3}/s) may be feasible with the prototype. However, additional extruder units (1{endash}3) would facilitate operations at the higher throughputs and reduce the duty cycle of each unit. The prototype can easily accommodate steady-state pellet fueling of present large tokamaks or other near-term plasma experiments.« less

Space reactor power 1986 - A year of choices and transition

NASA Technical Reports Server (NTRS)

Wiley, R. L.; Verga, R. L.; Schnyer, A. D.; Sholtis, J. A., Jr.; Wahlquist, E. J.

1986-01-01

Both the SP-100 and Multimegawatt programs have made significant progress over the last year and that progress is the focus of this paper. In the SP-100 program the thermoelectric energy conversion concept powered by a compact, high-temperature, lithium-cooled, uranium-nitride-fueled fast spectrum reactor was selected for engineering development and ground demonstration testing at an electrical power level of 300 kilowatts. In the Multimegawatt program, activities moved from the planning phase into one of technology development and assessment with attendant preliminary definition and evaluation of power concepts against requirements of the Strategic Defense Initiative.

Simulation of High-Beta Plasma Confinement

NASA Astrophysics Data System (ADS)

Font, Gabriel; Welch, Dale; Mitchell, Robert; McGuire, Thomas

2017-10-01

The Lockheed Martin Compact Fusion Reactor concept utilizes magnetic cusps to confine the plasma. In order to minimize losses through the axial and ring cusps, the plasma is pushed to a high-beta state. Simulations were made of the plasma and magnetic field system in an effort to quantify particle confinement times and plasma behavior characteristics. Computations are carried out with LSP using implicit PIC methods. Simulations of different sub-scale geometries at high-Beta fusion conditions are used to determine particle loss scaling with reactor size, plasma conditions, and gyro radii. ©2017 Lockheed Martin Corporation. All Rights Reserved.

Dual-phase reactor plant with partitioned isolation condenser

DOEpatents

Hui, Marvin M.

1992-01-01

A nuclear energy plant housing a boiling-water reactor utilizes an isolation condenser in which a single chamber is partitioned into a distributor plenum and a collector plenum. Steam accumulates in the distributor plenum and is conveyed to the collector plenum through an annular manifold that includes tubes extending through a condenser pool. The tubes provide for a transfer of heat from the steam, forming a condensate. The chamber has a disk-shaped base, a cylindrical sidewall, and a semispherical top. This geometry results in a compact design that exhibits significant performance and cost advantages over prior designs.

A strategy for intensive production of molybdenum-99 isotopes for nuclear medicine using CANDU reactors.

PubMed

Morreale, A C; Novog, D R; Luxat, J C

2012-01-01

Technetium-99m is an important medical isotope utilized worldwide in nuclear medicine and is produced from the decay of its parent isotope, molybdenum-99. The online fueling capability and compact fuel of the CANDU(®)(1) reactor allows for the potential production of large quantities of (99)Mo. This paper proposes (99)Mo production strategies using modified target fuel bundles loaded into CANDU fuel channels. Using a small group of channels a yield of 89-113% of the weekly world demand for (99)Mo can be obtained. Copyright © 2011 Elsevier Ltd. All rights reserved.

ASME code considerations for the compact heat exchanger

DOE Office of Scientific and Technical Information (OSTI.GOV)

Nestell, James; Sham, Sam

2015-08-31

The mission of the U.S. Department of Energy (DOE), Office of Nuclear Energy is to advance nuclear power in order to meet the nation's energy, environmental, and energy security needs. Advanced high temperature reactor systems such as sodium fast reactors and high and very high temperature gas-cooled reactors are being considered for the next generation of nuclear reactor plant designs. The coolants for these high temperature reactor systems include liquid sodium and helium gas. Supercritical carbon dioxide (sCO₂), a fluid at a temperature and pressure above the supercritical point of CO₂, is currently being investigated by DOE as a workingmore » fluid for a nuclear or fossil-heated recompression closed Brayton cycle energy conversion system that operates at 550°C (1022°F) at 200 bar (2900 psi). Higher operating temperatures are envisioned in future developments. All of these design concepts require a highly effective heat exchanger that transfers heat from the nuclear or chemical reactor to the chemical process fluid or the to the power cycle. In the nuclear designs described above, heat is transferred from the primary to the secondary loop via an intermediate heat exchanger (IHX) and then from the intermediate loop to either a working process or a power cycle via a secondary heat exchanger (SHX). The IHX is a component in the primary coolant loop which will be classified as "safety related." The intermediate loop will likely be classified as "not safety related but important to safety." These safety classifications have a direct bearing on heat exchanger design approaches for the IHX and SHX. The very high temperatures being considered for the VHTR will require the use of very high temperature alloys for the IHX and SHX. Material cost considerations alone will dictate that the IHX and SHX be highly effective; that is, provide high heat transfer area in a small volume. This feature must be accompanied by low pressure drop and mechanical reliability and robustness. Classic shell and tube designs will be large and costly, and may only be appropriate in steam generator service in the SHX where boiling inside the tubes occurs. For other energy conversion systems, all of these features can be met in a compact heat exchanger design. This report will examine some of the ASME Code issues that will need to be addressed to allow use of a Code-qualified compact heat exchanger in IHX or SHX nuclear service. Most effort will focus on the IHX, since the safety-related (Class A) design rules are more extensive than those for important-to-safety (Class B) or commercial rules that are relevant to the SHX.« less

Various methods to improve heat transfer in exchangers

NASA Astrophysics Data System (ADS)

Pavel, Zitek; Vaclav, Valenta

2015-05-01

The University of West Bohemia in Pilsen (Department of Power System Engineering) is working on the selection of effective heat exchangers. Conventional shell and tube heat exchangers use simple segmental baffles. It can be replaced by helical baffles, which increase the heat transfer efficiency and reduce pressure losses. Their usage is demonstrated in the primary circuit of IV. generation MSR (Molten Salt Reactors). For high-temperature reactors we consider the use of compact desk heat exchangers, which are small, which allows the integral configuration of reactor. We design them from graphite composites, which allow up to 1000°C and are usable as exchangers: salt-salt or salt-acid (e.g. for the hydrogen production). In the paper there are shown thermo-physical properties of salts, material properties and principles of calculations.

Surface morphology changes to tungsten under exposure to He ions from an electron cyclotron resonance plasma source

NASA Astrophysics Data System (ADS)

Donovan, David; Maan, Anurag; Duran, Jonah; Buchenauer, Dean; Whaley, Josh

2015-11-01

Exposure of tungsten to low energy (<100 eV) helium plasmas at temperatures between 900-1900 K in both laboratory experiments and tokamaks has been shown to cause severe nanoscale modification of the near surface resulting the growth of tungsten tendrils. We used a relatively low flux (2.5x1019 ions m-2 s-1) compact ECR plasma source at Sandia-California to investigate the early stages of helium induced tungsten damage. Exposures of polished tungsten discs were performed and characterized using SEM, AFM, and FIB cross section imaging. Bubbles have been seen on the exposed tungsten surface and in sub-surface cross sections growing to up to 150 nm in diameter. Comparisons were made between exposures of warm rolled Plansee tungsten discs and ALMT ITER grade tungsten samples. A similar He plasma exposure stage has now been developed at the University of Tennessee-Knoxville with an improved compact ECR plasma source. Status of the new UTK exposure stage will be discussed as well as planned experiments and new material characterization techniques (EBSD, GIXRD). Work supported by US DOE Contract DE-AC04-94AL85000 and the PSI Science Center.

DOE Office of Scientific and Technical Information (OSTI.GOV)

Batchelor, D.B.; Carreras, B.A.; Hirshman, S.P.

Significant progress has been made in the development of new modest-size compact stellarator devices that could test optimization principles for the design of a more attractive reactor. These are 3 and 4 field period low-aspect-ratio quasi-omnigenous (QO) stellarators based on an optimization method that targets improved confinement, stability, ease of coil design, low-aspect-ratio, and low bootstrap current.

CONTROL ROD

DOEpatents

Walker, D.E.; Matras, S.

1963-04-30

This patent shows a method of making a fuel or control rod for a nuclear reactor. Fuel or control material is placed within a tube and plugs of porous metal wool are inserted at both ends. The metal wool is then compacted and the tube compressed around it as by swaging, thereby making the plugs liquid- impervious but gas-pervious. (AEC)

Runaway Geneeration In Disruptions Of Plasmas In TFTR

DOE Office of Scientific and Technical Information (OSTI.GOV)

Fredrickson, E. D.; Bell, M. G.; Taylor, G.

2014-03-31

Many disruptions in the Tokamak Fusion Test Reactor (TFTR) [D. Meade and the TFTR Group, in Proceedings of the International Conference on Plasma Physics and Controlled Nuclear Fusion, Washington, DC, 1990 (International Atomic Energy Agency, Vienna, 1991), Vol. 1, pp. 9-24] produced populations of runaway electrons which carried a significant fraction of the original plasma current. In this paper, we describe experiments where, following a disruption of a low-beta, reversed shear plasma, currents of up to 1 MA carried mainly by runaway electrons were controlled and then ramped down to near zero using the ohmic transformer. In the longer lastingmore » runaway plasmas, Parail-Pogutse instabilities were observed.« less

Hardwired Control Changes For NSTX DC Power Feeds

DOE Office of Scientific and Technical Information (OSTI.GOV)

Ramakrishnan, S.

The National Spherical Torus Experiment (NSTX) has been designed and installed in the existing facilities at Princeton Plasma Physics Laboratory (PPPL). Most of the hardware, plant facilities, auxiliary sub-systems, and power systems originally used for the Tokamak Fusion Test Reactor (TFTR) have been used with suitable modifications to reflect NSTX needs. The original TFTR Hardwired Control System (HCS) with electromechanical relays was used for NSTX DC Power loop control and protection during NSTX operations. As part of the NSTX Upgrade, the HCS is being changed to a PLC-based system with the same control logic. This paper gives a description ofmore » the changeover to the new PLC-based system __________________________________________________« less

MM-wave cyclotron auto-resonance maser for plasma heating

NASA Astrophysics Data System (ADS)

Ceccuzzi, S.; Dattoli, G.; Di Palma, E.; Doria, A.; Gallerano, G. P.; Giovenale, E.; Mirizzi, F.; Spassovsky, I.; Ravera, G. L.; Surrenti, V.; Tuccillo, A. A.

2014-02-01

Heating and Current Drive systems are of outstanding relevance in fusion plasmas, magnetically confined in tokamak devices, as they provide the tools to reach, sustain and control burning conditions. Heating systems based on the electron cyclotron resonance (ECRH) have been extensively exploited on past and present machines DEMO, and the future reactor will require high frequencies. Therefore, high power (≥1MW) RF sources with output frequency in the 200 - 300 GHz range would be necessary. A promising source is the so called Cyclotron Auto-Resonance Maser (CARM). Preliminary results of the conceptual design of a CARM device for plasma heating, carried out at ENEA-Frascati will be presented together with the planned R&D development.

Alpha effect of Alfv{acute e}n waves and current drive in reversed-field pinches

DOE Office of Scientific and Technical Information (OSTI.GOV)

Litwin, C.; Prager, S.C.

Circularly polarized Alfv{acute e}n waves give rise to an {alpha}-dynamo effect that can be exploited to drive parallel current. In a {open_quotes}laminar{close_quotes} magnetic the effect is weak and does not give rise to significant currents for realistic parameters (e.g., in tokamaks). However, in reversed-field pinches (RFPs) in which magnetic field in the plasma core is stochastic, a significant enhancement of the {alpha} effect occurs. Estimates of this effect show that it may be a realistic method of current generation in the present-day RFP experiments and possibly also in future RFP-based fusion reactors. {copyright} {ital 1998 American Institute of Physics.}

ITER CS Model Coil and CS Insert Test Results

DOE Office of Scientific and Technical Information (OSTI.GOV)

Martovetsky, N; Michael, P; Minervina, J

2000-09-07

The Inner and Outer modules of the Central Solenoid Model Coil (CSMC) were built by US and Japanese home teams in collaboration with European and Russian teams to demonstrate the feasibility of a superconducting Central Solenoid for ITER and other large tokamak reactors. The CSMC mass is about 120 t, OD is about 3.6 m and the stored energy is 640 MJ at 46 kA and peak field of 13 T. Testing of the CSMC and the CS Insert took place at Japan Atomic Energy Research Institute (JAERI) from mid March until mid August 2000. This paper presents the mainmore » results of the tests performed.« less

Gyrotron-driven high current ECR ion source for boron-neutron capture therapy neutron generator

NASA Astrophysics Data System (ADS)

Skalyga, V.; Izotov, I.; Golubev, S.; Razin, S.; Sidorov, A.; Maslennikova, A.; Volovecky, A.; Kalvas, T.; Koivisto, H.; Tarvainen, O.

2014-12-01

Boron-neutron capture therapy (BNCT) is a perspective treatment method for radiation resistant tumors. Unfortunately its development is strongly held back by a several physical and medical problems. Neutron sources for BNCT currently are limited to nuclear reactors and accelerators. For wide spread of BNCT investigations more compact and cheap neutron source would be much more preferable. In present paper an approach for compact D-D neutron generator creation based on a high current ECR ion source is suggested. Results on dense proton beams production are presented. A possibility of ion beams formation with current density up to 600 mA/cm2 is demonstrated. Estimations based on obtained experimental results show that neutron target bombarded by such deuteron beams would theoretically yield a neutron flux density up to 6·1010 cm-2/s. Thus, neutron generator based on a high-current deuteron ECR source with a powerful plasma heating by gyrotron radiation could fulfill the BNCT requirements significantly lower price, smaller size and ease of operation in comparison with existing reactors and accelerators.

HYDRAULIC SERVO CONTROL MECHANISM

DOEpatents

Hussey, R.B.; Gottsche, M.J. Jr.

1963-09-17

A hydraulic servo control mechanism of compact construction and low fluid requirements is described. The mechanism consists of a main hydraulic piston, comprising the drive output, which is connected mechanically for feedback purposes to a servo control piston. A control sleeve having control slots for the system encloses the servo piston, which acts to cover or uncover the slots as a means of controlling the operation of the system. This operation permits only a small amount of fluid to regulate the operation of the mechanism, which, as a result, is compact and relatively light. This mechanism is particuiarly adaptable to the drive and control of control rods in nuclear reactors. (auth)

METHOD OF MAKING FUEL BODIES

DOEpatents

Goeddel, W.V.; Simnad, M.T.

1962-04-24

An improved method of making a fuel body containing carbon for reactors is described. Carbides of uranium and thorium having a particle size of from 100 to 500 microns are mixed with carbon having a particle size that will pass a 200 mesh screen but be retained by a 325 mesh screen, and 10 per cent by weight pitch. The mixture is heated to a temperature of about 700 to 900 deg C, at which point bonding is effected while maintaining it under mechanical pressure of over 3,000 pounds per square inch. The entire compact is heated to a uniform temperature during the process, preferably by electrical resistance of the compact itself. (AEC)

Fuel burnup analysis for IRIS reactor using MCNPX and WIMS-D5 codes

NASA Astrophysics Data System (ADS)

Amin, E. A.; Bashter, I. I.; Hassan, Nabil M.; Mustafa, S. S.

2017-02-01

International Reactor Innovative and Secure (IRIS) reactor is a compact power reactor designed with especial features. It contains Integral Fuel Burnable Absorber (IFBA). The core is heterogeneous both axially and radially. This work provides the full core burn up analysis for IRIS reactor using MCNPX and WIMDS-D5 codes. Criticality calculations, radial and axial power distributions and nuclear peaking factor at the different stages of burnup were studied. Effective multiplication factor values for the core were estimated by coupling MCNPX code with WIMS-D5 code and compared with SAS2H/KENO-V code values at different stages of burnup. The two calculation codes show good agreement and correlation. The values of radial and axial powers for the full core were also compared with published results given by SAS2H/KENO-V code (at the beginning and end of reactor operation). The behavior of both radial and axial power distribution is quiet similar to the other data published by SAS2H/KENO-V code. The peaking factor values estimated in the present work are close to its values calculated by SAS2H/KENO-V code.

Local Helicity Injection Systems for Non-solenoidal Startup in the PEGASUS Toroidal Experiment

NASA Astrophysics Data System (ADS)

Perry, J. M.; Barr, J. L.; Bongard, M. W.; Fonck, R. J.; Hinson, E. T.; Lewicki, B. T.; Redd, A. J.

2013-10-01

Local helicity injection is being developed in the PEGASUS Toroidal Experiment for non-solenoidal startup in spherical tokamaks. The effective loop voltage due to helicity injection scales with the area of the injectors, requiring the development of electron current injectors with areas much larger than the 2 cm2 plasma arc injectors used to date. Solid and gas-effused metallic electrodes were found to be unusable due to reduced injector area utilization from localized cathode spots and narrow operational regimes. An integrated array of 8 compact plasma arc sources is thus being developed for high current startup. It employs two monolithic power systems, for the plasma arc sources and the bias current extraction system. The array effectively eliminates impurity fueling from plasma-material interaction by incorporating a local scraper-limiter and conical-frustum bias electrodes to mitigate the effects of cathode spots. An energy balance model of helicity injection indicates that the resulting 20 cm2 of total injection area should provide sufficient current drive to reach 0.3 MA. At that level, helicity injection drive exceeds that from poloidal induction, which is the relevant operational regime for large-scale spherical tokamaks. Future placement of the injector array near an expanded boundary divertor region will test simultaneous optimization of helicity drive and the Taylor relaxation current limit. Work supported by US DOE Grant DE-FG02-96ER54375.

Compact and Lightweight Sabatier Reactor for Carbon Dioxide Reduction

NASA Technical Reports Server (NTRS)

Junaedi, Christian; Hawley, Kyle; Walsh, Dennis; Roychoudhury, Subir; Abney, Morgan B.; Perry, Jay L.

2011-01-01

The utilization of CO2 to produce life support consumables, such as O2 and H2O, via the Sabatier reaction is an important aspect of NASA s cabin Atmosphere Revitalization System and In-Situ Resource Utilization architectures for both low-earth orbit and long-term manned space missions. In the current International Space Station (ISS) and other low orbit missions, metabolically-generated CO2 is removed from the cabin air and vented into space, resulting in a net loss of O2. This requires a continuous resupply of O2 via water electrolysis, and thus highlights the need for large water storage capacity. For long-duration space missions, the amount of life support consumables is limited and resupply options are practically nonexistent, thus atmosphere resource management and recycle becomes crucial to significantly reduce necessary O2 and H2O storage. Additionally, the potential use of the Martian CO2-rich atmosphere and Lunar regolith to generate life support consumables and propellant fuels is of interest to NASA. Precision Combustion, Inc. (PCI) has developed a compact, lightweight Microlith(Registered TradeMark)-based Sabatier (CO2 methanation) reactor which demonstrates the capability of achieving high CO2 conversion and near 100% CH4 selectivity at space velocities of 30,000-60,000 hr-1. The combination of the Microlith(Registered TradeMark) substrates and durable, novel catalyst coating permitted efficient Sabatier reactor operation that favors high reactant conversion, high selectivity, and long-term durability. This paper presents the reactor development and performance results at various operating conditions. Additionally, results from 100-hr durability tests and mechanical vibration tests are discussed.

The UASB reactor as an alternative for the septic tank for on-site sewage treatment.

PubMed

Coelho, A L S S; do Nascimento, M B H; Cavalcanti, P F F; van Haandel, A C

2003-01-01

Although septic tanks are amply used for on site sewage treatment, these units have serious drawbacks: the removal efficiency of organic material and suspended solids is low, the units are costly and occupy a large area and operational cost is high due to the need for periodic desludging. In this paper an innovative variant of the UASB reactor is proposed as an alternative for the septic tank. This alternative has several important advantages in comparison with the conventional septic tank: (1) Although the volume of the UASB reactor was about 4 times smaller than the septic tank, its effluent quality was superior, even though small sludge particles were present, (2) desludging of the UASB reactor is unnecessary and even counterproductive, as the sludge mass guarantees proper performance, (3) the UASB reactor is easily transportable (compact and light) and therefore can be produced in series, strongly reducing construction costs and (4) since the concentration of colloids in the UASB effluent is much smaller than in the ST effluent, it is expected that the infiltration of the effluent will be much less problematic.

Automated Identification of MHD Mode Bifurcation and Locking in Tokamaks

NASA Astrophysics Data System (ADS)

Riquezes, J. D.; Sabbagh, S. A.; Park, Y. S.; Bell, R. E.; Morton, L. A.

2017-10-01

Disruption avoidance is critical in reactor-scale tokamaks such as ITER to maintain steady plasma operation and avoid damage to device components. A key physical event chain that leads to disruptions is the appearance of rotating MHD modes, their slowing by resonant field drag mechanisms, and their locking. An algorithm has been developed that automatically detects bifurcation of the mode toroidal rotation frequency due to loss of torque balance under resonant braking, and mode locking for a set of shots using spectral decomposition. The present research examines data from NSTX, NSTX-U and KSTAR plasmas which differ significantly in aspect ratio (ranging from A = 1.3 - 3.5). The research aims to examine and compare the effectiveness of different algorithms for toroidal mode number discrimination, such as phase matching and singular value decomposition approaches, and to examine potential differences related to machine aspect ratio (e.g. mode eigenfunction shape variation). Simple theoretical models will be compared to the dynamics found. Main goals are to detect or potentially forecast the event chain early during a discharge. This would serve as a cue to engage active mode control or a controlled plasma shutdown. Supported by US DOE Contracts DE-SC0016614 and DE-AC02-09CH11466.

Numerical Study of High Heat Flux Performances of Flat-Tile Divertor Mock-ups with Hypervapotron Cooling Concept

NASA Astrophysics Data System (ADS)

Chen, Lei; Liu, Xiang; Lian, Youyun; Cai, Laizhong

2015-09-01

The hypervapotron (HV), as an enhanced heat transfer technique, will be used for ITER divertor components in the dome region as well as the enhanced heat flux first wall panels. W-Cu brazing technology has been developed at SWIP (Southwestern Institute of Physics), and one W/CuCrZr/316LN component of 450 mm×52 mm×166 mm with HV cooling channels will be fabricated for high heat flux (HHF) tests. Before that a relevant analysis was carried out to optimize the structure of divertor component elements. ANSYS-CFX was used in CFD analysis and ABAQUS was adopted for thermal-mechanical calculations. Commercial code FE-SAFE was adopted to compute the fatigue life of the component. The tile size, thickness of tungsten tiles and the slit width among tungsten tiles were optimized and its HHF performances under International Thermonuclear Experimental Reactor (ITER) loading conditions were simulated. One brand new tokamak HL-2M with advanced divertor configuration is under construction in SWIP, where ITER-like flat-tile divertor components are adopted. This optimized design is expected to supply valuable data for HL-2M tokamak. supported by the National Magnetic Confinement Fusion Science Program of China (Nos. 2011GB110001 and 2011GB110004)

Control of three dimensional particle flux to divertor using rotating RMP in the EAST tokamak

NASA Astrophysics Data System (ADS)

Jia, M.; Sun, Y.; Liang, Y.; Wang, L.; Xu, J.; Gu, S.; Lyu, B.; Wang, H. H.; Yang, X.; Zhong, F.; Chu, N.; Feng, W.; He, K.; Liu, Y. Q.; Qian, J.; Shi, T.; Shen, B.

2018-04-01

Controlling the steady state particle and heat flux impinging on the plasma facing components, as one of the main concerns of future fusion reactors, is still necessary when the transient power loads induced by edge localized modes (ELMs) have been eliminated by resonant magnetic perturbations (RMPs) in high confinement tokamak experiments. This is especially true for long pulse operation. One promising solution is to use the rotating perturbed field. Recently rotating and differential phase scans of n  =  1 and 2 RMP fields have been operated for the first time in EAST discharges. The particle flux patterns on the divertor targets change synchronously with both rotating and phasing RMP fields as predicted by the modeled magnetic footprint patterns. The modeling with plasma response, which is calculated by MARS-F, is also carried out. The plasma response shows amplifying or screening effect to n  =  2 perturbations with different spectra. This changes the field line penetration depth rather than the general footprint shape. This has been verified by experimental observations on EAST. These experiments motivate further study of reducing both transient and steady state local power load and particle flux with the help of rotating RMPs in long pulse operation.

Helium, Iron and Electron Particle Transport and Energy Transport Studies on the TFTR Tokamak

DOE R&D Accomplishments Database

Synakowski, E. J.; Efthimion, P. C.; Rewoldt, G.; Stratton, B. C.; Tang, W. M.; Grek, B.; Hill, K. W.; Hulse, R. A.; Johnson, D .W.; Mansfield, D. K.; McCune, D.; Mikkelsen, D. R.; Park, H. K.; Ramsey, A. T.; Redi, M. H.; Scott, S. D.; Taylor, G.; Timberlake, J.; Zarnstorff, M. C. (Princeton Univ., NJ (United States). Plasma Physics Lab.); Kissick, M. W. (Wisconsin Univ., Madison, WI (United States))

1993-03-01

Results from helium, iron, and electron transport on TFTR in L-mode and Supershot deuterium plasmas with the same toroidal field, plasma current, and neutral beam heating power are presented. They are compared to results from thermal transport analysis based on power balance. Particle diffusivities and thermal conductivities are radially hollow and larger than neoclassical values, except possibly near the magnetic axis. The ion channel dominates over the electron channel in both particle and thermal diffusion. A peaked helium profile, supported by inward convection that is stronger than predicted by neoclassical theory, is measured in the Supershot The helium profile shape is consistent with predictions from quasilinear electrostatic drift-wave theory. While the perturbative particle diffusion coefficients of all three species are similar in the Supershot, differences are found in the L-Mode. Quasilinear theory calculations of the ratios of impurity diffusivities are in good accord with measurements. Theory estimates indicate that the ion heat flux should be larger than the electron heat flux, consistent with power balance analysis. However, theoretical values of the ratio of the ion to electron heat flux can be more than a factor of three larger than experimental values. A correlation between helium diffusion and ion thermal transport is observed and has favorable implications for sustained ignition of a tokamak fusion reactor.

Snowflake divertor experiments in the DIII-D, NSTX, and NSTX-U tokamaks aimed at the development of the divertor power exhaust solution

DOE PAGES

Soukhanovskii, V. A.; Allen, S. L.; Fenstermacher, M. E.; ...

2016-11-16

Experimental results from the National Spherical Torus Experiment (NSTX), a medium-size spherical tokamak with a compact divertor, and DIII-D, a large conventional aspect ratio tokamak, demonstrate that the snowflake (SF) divertor configuration may provide a promising solution for mitigating divertor heat loads and target plate erosion compatible with core H-mode confinement in the future fusion devices, where the standard radiative divertor solution may be inadequate. In NSTX, where the initial high-power SF experiment was performed, the SF divertor was compatible with H-mode confinement, and led to the destabilization of large Edge Localized Modes (ELMs). However, a stable partial detachment ofmore » the outer strike point was also achieved where inter-ELM peak heat flux was reduced by factors 3-5, and peak ELM heat flux was reduced by up to 80% (see standard divertor). The DIII-D studies show the SF divertor enables significant power spreading in attached and radiative divertor conditions. Results include: compatibility with the core and pedestal, peak inter-ELM divertor heat flux reduction due to geometry at lower ne, and ELM energy and divertor peak heat flux reduction, especially prominent in radiative D 2-seeded SF divertor, and nearly complete power detachment and broader radiated power distribution in the radiative D 2-seeded SF divertor at PSOL = 3 - 4 MW. A variety of SF configurations can be supported by the divertor coil set in NSTX Upgrade. Edge transport modeling with the multifluid edge transport code UEDGE shows that the radiative SF divertor can successfully reduce peak divertor heat flux for the projected PSOL ≃ 9 MW case. Furthermore, the radiative SF divertor with carbon impurity provides a wider ne operating window, 50% less argon is needed in the impurity-seeded SF configuration to achieve similar q peak reduction factors (see standard divertor).« less

The dependence of divertor power sharing on magnetic flux balance in near double-null configurations on Alcator C-Mod

NASA Astrophysics Data System (ADS)

Brunner, D.; Kuang, A. Q.; LaBombard, B.; Terry, J. L.

2018-07-01

Management of power exhaust will be a crucial task for tokamak fusion reactors. Reactor concepts are often proposed with double-null divertors, i.e. having two magnetic separatrices in an up-down symmetric configuration. This arrangement is potentially advantageous since the majority of the tokamak exhaust power tends to flow to the outer pair of divertor legs at large major radius, where the geometry is favorable for spreading the heat over a large surface area and there is more room for advanced divertor configurations. Despite the importance, there have been relatively few studies of divertor power sharing in near double null configurations and no studies at the poloidal magnetic fields and scrape-off layer power widths anticipated for a reactor. Motivated by this need we have undertaken a systematic study on Alcator C-Mod, examining the effect of magnetic flux balance on the power sharing among the four divertor legs in near double-null plasmas. Ohmic L-modes at three values of plasma current and ICRF-heated enhanced D-alpha (EDA) H-modes and I-modes at a single value of plasma current are explored, producing poloidal magnetic fields of 0.42, 0.62 and 0.85 Tesla. For Ohmic L-modes and ICRF-heated EDA H-modes, we find that the point of equal power sharing between upper and lower divertors occurs remarkably close to a balanced double null. Power sharing amongst the outer (upper versus lower) and inner (upper versus lower) pairs of divertors can be described in terms of a logistic function of magnetic flux balance, consistent with heat flux mapping along magnetic field lines to the outer midplane. Power sharing between inner and outer legs is found to follow a Gaussian-like function of magnetic flux balance with non-zero power to the inner divertors at double null. The overall behavior of H-modes operated near double null and for I-modes operating to within one heat flux e-folding of double null are found similar to Ohmic L-modes, with a significant reduction of power on the inner divertor legs. The results are encapsulated in terms of empirically-informed analytic functions of magnetic flux balance. When combined with magnetic equilibrium control system specifications, these relationships can be used to specify the power flux handling requirements for each of the four divertor target plates.

Fusion Plasma Performance and Confinement Studies on JT-60 and JT-60U

DOE Office of Scientific and Technical Information (OSTI.GOV)

Kamada, Y.; Fujita, T.; Ishida, S.

2002-09-15

Fusion plasma performance and confinement studies on JT-60 and JT-60U are reviewed. With the main aim of providing a physics basis for ITER and the steady-state tokamak reactors, JT-60/JT-60U has been developing and optimizing the operational concepts, and extending the discharge regimes toward sustainment of high integrated performance in the reactor relevant parameter regime. In addition to achievement of high fusion plasma performances such as the equivalent breakeven condition (Q{sub DT}{sup eq} up to 1.25) and a high fusion triple product n{sub D}(0){tau}{sub E}T{sub i}(0) = 1.5 x 10{sup 21} m{sup -3}skeV, JT-60U has demonstrated the integrated performance of highmore » confinement, high {beta}{sub N}, full non-inductive current drive with a large fraction of bootstrap current. These favorable performances have been achieved in the two advanced operation regimes, the reversed magnetic shear (RS) and the weak magnetic shear (high-{beta}{sub p}) ELMy H modes characterized by both internal transport barriers (ITB) and edge transport barriers (ETB). The key factors in optimizing these plasmas towards high integrated performance are control of profiles of current, pressure, rotation, etc. utilizing a variety of heating, current drive, torque input, and particle control capabilities and high triangularity operation. As represented by discovery of ITBs (density ITB in the central pellet mode, ion temperature ITB in the high-{beta}{sub p} mode, and electron temperature ITB in the reversed shear mode), confinement studies in JT-60/JT-60U have been emphasizing freedom and also restriction of radial profiles of temperature and density. In addition to characterization of confinement and analyses of transport properties of the OH, the L-mode, the H-mode, the pellet mode, the high-{beta}{sub p} mode, and the RS mode, JT-60U has clarified formation conditions, spatial structures and dynamics of edge and internal transport barriers, and evaluated effects of repetitive MHD events on confinement such as sawteeth and ELMs. Through these studies, JT-60U has demonstrated applicability of the high confinement modes to ITER and the steady-state tokamak reactors.« less

Reactor antineutrino detector iDREAM.

NASA Astrophysics Data System (ADS)

Gromov, M. B.; Lukyanchenko, G. A.; Novikova, G. J.; Obinyakov, B. A.; Oralbaev, A. Y.; Skorokhvatov, M. D.; Sukhotin, S. V.; Chepurnov, A. S.; Etenko, A. V.

2017-09-01

Industrial Detector for Reactor Antineutrino Monitoring (iDREAM) is a compact (≈ 3.5m 2) industrial electron antineutrino spectrometer. It is dedicated for remote monitoring of PWR reactor operational modes by neutrino method in real-time. Measurements of antineutrino flux from PWR allow to estimate a fuel mixture in active zone and to check the status of the reactor campaign for non-proliferation purposes. LAB-based gadolinium doped scintillator is exploited as a target. Multizone architecture of the detector with gamma-catcher surrounding fiducial volume and plastic muon veto above and below ensure high efficiency of IBD detection and background suppression. DAQ is based on Flash ADC with PSD discrimination algorithms while digital trigger is programmable and flexible due to FPGA. The prototype detector was started up in 2014. Preliminary works on registration Cerenkov radiation produced by cosmic muons were established with distilled water inside the detector in order to test electronic and slow control systems. Also in parallel a long-term measurements with different scintillator samples were conducted.

FALCON nuclear-reactor-pumped laser program and wireless power transmission

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, R.J.; Pickard, P.S.

1992-12-31

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

FALCON nuclear-reactor-pumped laser program and wireless power transmission

DOE Office of Scientific and Technical Information (OSTI.GOV)

Lipinski, R.J.; Pickard, P.S.

1992-01-01

FALCON is a high-power, reactor-pumped laser concept. The major strengths of a reactor-pumped laser are (1) simple, modular construction, (2) long-duration, closed-cycle capability, (3) self-contained power, (4) compact size, and (5) a variety of wavelengths (from visible to infrared). Reactor-pumped lasing has been demonstrated experimentally in various mixtures of xenon, argon, neon, and helium at wavelengths of 585, 703, 725, 1271, 1733, 1792, 2032, 2630, 2650, and 3370 nm with intrinsic efficiency as high as 2.5%. Powers up to 300 W for 2 ms have been demonstrated. Projected beam quality for FALCON is good enough that frequency doubling at reasonablemore » efficiencies could be expected to yield wavelengths at 353, 363, 636, 867, 896, 1016, 1315, 1325, and 1685 nm. Appropriate missions for FALCON are described and include power beaming to satellites, the moon, and unmanned surveillance planes; lunar mapping; space debris removal; and laser propulsion.« less

Addition of acetate improves stability of power generation using microbial fuel cells treating domestic wastewater.

PubMed

Stager, Jennifer L; Zhang, Xiaoyuan; Logan, Bruce E

2017-12-01

Power generation using microbial fuel cells (MFCs) must provide stable, continuous conversion of organic matter in wastewaters into electricity. However, when relatively small diameter (0.8cm) graphite fiber brush anodes were placed close to the cathodes in MFCs, power generation was unstable during treatment of low strength domestic wastewater. One reactor produced 149mW/m 2 before power generation failed, while the other reactor produced 257mW/m 2 , with both reactors exhibiting severe power overshoot in polarization tests. Using separators or activated carbon cathodes did not result in stable operation as the reactors continued to exhibit power overshoot based on polarization tests. However, adding acetate (1g/L) to the wastewater produced stable performance during fed batch and continuous flow operation, and there was no power overshoot in polarization tests. These results highlight the importance of wastewater strength and brush anode size for producing stable and continuous power in compact MFCs. Copyright © 2017 Elsevier B.V. All rights reserved.

An MFC-Based Online Monitoring and Alert System for Activated Sludge Process

PubMed Central

Xu, Gui-Hua; Wang, Yun-Kun; Sheng, Guo-Ping; Mu, Yang; Yu, Han-Qing

2014-01-01

In this study, based on a simple, compact and submersible microbial fuel cell (MFC), a novel online monitoring and alert system with self-diagnosis function was established for the activated sludge (AS) process. Such a submersible MFC utilized organic substrates and oxygen in the AS reactor as the electron donor and acceptor respectively, and could provide an evaluation on the status of the AS reactor and thus give a reliable early warning of potential risks. In order to evaluate the reliability and sensitivity of this online monitoring and alert system, a series of tests were conducted to examine the response of this system to various shocks imposed on the AS reactor. The results indicate that this online monitoring and alert system was highly sensitive to the performance variations of the AS reactor. The stability, sensitivity and repeatability of this online system provide feasibility of being incorporated into current control systems of wastewater treatment plants to real-time monitor, diagnose, alert and control the AS process. PMID:25345502

A new simpler way to obtain high fusion power gain in tandem mirrors

NASA Astrophysics Data System (ADS)

Fowler, T. K.; Moir, R. W.; Simonen, T. C.

2017-05-01

From the earliest days of fusion research, Richard F. Post and other advocates of magnetic mirror confinement recognized that mirrors favor high ion temperatures where nuclear reaction rates < σ v> begin to peak for all fusion fuels. In this paper we review why high ion temperatures are favored, using Post’s axisymmetric Kinetically Stabilized Tandem Mirror as the example; and we offer a new idea that appears to greatly improve reactor prospects at high ion temperatures. The idea is, first, to take advantage of recent advances in superconducting magnet technology to minimize the size and cost of End Plugs; and secondly, to utilize parallel advances in gyrotrons that would enable intense electron cyclotron heating (ECH) in these high field End Plugs. The yin-yang magnets and thermal barriers that complicated earlier tandem mirror designs are not required. We find that, concerning end losses, intense ECH in symmetric End Plugs could increase the fusion power gain Q, for both DT and Catalyzed DD fuel cycles, to levels competitive with steady-state tokamaks burning DT fuel. Radial losses remain an issue that will ultimately determine reactor viability.

Effect of the self-pumped limiter concept on the tritium fuel cycle

DOE Office of Scientific and Technical Information (OSTI.GOV)

Finn, P.A.; Sze, D.K.; Hassanein, A.

1988-09-01

The self-pumped limiter concept was the impurity control system for the reactor in the Tokamak Power Systems Study (TPSS). The use of a self-pumped limiter had a major impact on the design of the tritium systems of the TPSS fusion reactor. The self-pumped limiter functions by depositing the helium ash under a layer of metal (vanadium). The majority of the hydrogen species are recycled at the plasma edge; a small fraction permeates to the blanket/coolant which was lithium in TPSS. Use of the self-pumped limiter results in the elimination of the plasma processing system. Thus, the blanket tritium processing systemmore » becomes the major tritium system. The main advantages achieved for the tritium systems with a self-pumped limiter are a reduction in the capital cost of tritium processing equipment as well as a reduction in building space, a reduced tritium inventory for processing and for reserve storage, an increase in the inherent safety of the fusion plant and an improvement in economics for a fusion world economy.« less

PROGRESS IN DESIGN OF THE INSTRUMENTATION AND CONTROL OF THE TOKAMAK COOLING WATER SYSTEM

DOE Office of Scientific and Technical Information (OSTI.GOV)

Korsah, Kofi; DeVan, Bill; Ashburn, David

This paper discusses progress in the design of the control, interlock and safety systems of the Tokamak Cooling Water System (TCWS) for the ITER fusion reactor. The TCWS instrumentation and control (I&C) is one of approximately 200 separate plant I&C systems (e.g., vacuum system I&C, magnets system I&C) that interface to a common central I&C system through standardized networks. Several aspects of the I&C are similar to the I&C of fission-based power plants. However, some of the unique features of the ITER fusion reactor and the TCWS (e.g., high quasi-static magnetic field, need for baking and drying as well asmore » cooling operations), also demand some unique safety and qualification considerations. The paper compares the design strategy/guidelines of the TCWS I&C and the I&C of conventional nuclear power plants. Issues such as safety classifications, independence between control and safety systems, sensor sharing, redundancy, voting schemes, and qualification methodologies are discussed. It is concluded that independence and separation requirements are similar in both designs. However, the voting schemes for safety systems in nuclear power plants typically use 2oo4 (i.e., 4 divisions of safety I&C, any 2 of which is sufficient to trigger a safety action), while 2oo3 voting logic - within each of 2 independent trains - is used in the TCWS I&C. It is also noted that 2oo3 voting is also acceptable in nuclear power plants if adequate risk assessment and reliability is demonstrated. Finally, while qualification requirements provide similar guidance [e.g., both IEC 60780 (invoked in ITER-space), and IEEE 323 (invoked in fission power plant space) provide similar guidance], an important qualification consideration is the susceptibility of I&C to the magnetic fields of ITER. Also, the radiation environments are different. In the case of magnetic fields the paper discusses some options that are being considered.« less

Fast start-up reactor for partial oxidation of methane with electrically-heated metallic monolith catalyst

NASA Astrophysics Data System (ADS)

Jung, Heon; Yoon, Wang Lai; Lee, Hotae; Park, Jong Soo; Shin, Jang Sik; La, Howon; Lee, Jong Dae

A palladium-washcoated metallic monolith catalyst is applied to the partial oxidation of methane to syngas. This catalyst is highly active at a gas hourly space velocity (GHSV) of 100,000 h -1. The compact partial oxidation (POX) reactor equipped with both 96 cc of the metallic monolith catalyst and an electrically-heated catalyst (EHC) has a start-up time of less than 1.5 min and a syngas generation capacity of 9.5 Nm 3 h -1. The POX reaction is sustained without the need for an external heater. With the stand-alone POX reactor, the methane conversion can be increased either by preheating the reactant mixture heat-exchanged with the product gas, or by supplying a larger amount of oxygen than is necessary for the reaction stoichiometry.

Appendix to HDC 2118 design criteria 100-X reactor water plant, general description - section II

DOE Office of Scientific and Technical Information (OSTI.GOV)

Not Available

1952-03-29

The factors responsible for the advances of 100-X compared with the older areas are: Simplification of the process, such as elimination of separate process water clearwells, by having the filtered water reservoirs perform that function. Combination of separate buildings into one building, such as combining filter pump house and process pump house. Use of electric standby. Use of higher capacity pumps and filter basins, and so fewer number of units. Centralization of control and operation. More compact arrangement of plant components. Use of waste heat for space heating, recovered from reactor effluent, backed up by steam plant.

Detection of on-surface objects with an underground radiography detector system using cosmic-ray muons

NASA Astrophysics Data System (ADS)

Fujii, Hirofumi; Hara, Kazuhiko; Hayashi, Kohei; Kakuno, Hidekazu; Kodama, Hideyo; Nagamine, Kanetada; Sato, Kazuyuki; Sato, Kotaro; Kim, Shin-Hong; Suzuki, Atsuto; Takahashi, Kazuki; Takasaki, Fumihiko

2017-05-01

We have developed a compact muon radiography detector to investigate the status of the nuclear debris in the Fukushima Daiichi Reactors. Our previous observation showed that a large portion of the Unit-1 Reactor fuel had fallen to floor level. The detector must be located underground to further investigate the status of the fallen debris. To investigate the performance of muon radiography in such a situation, we observed 2 m cubic iron blocks located on the surface of the ground through different lengths of ground soil. The iron blocks were imaged and their corresponding iron density was derived successfully.

Nuclear powerplants for mobile applications.

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. This paper examines the technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

Nuclear power plants for mobile applications

NASA Technical Reports Server (NTRS)

Anderson, J. L.

1972-01-01

Mobile nuclear powerplants for applications other than large ships and submarines will require compact, lightweight reactors with especially stringent impact-safety design. The technical and economic feasibility that the broadening role of civilian nuclear power, in general, (land-based nuclear electric generating plants and nuclear ships) can extend to lightweight, safe mobile nuclear powerplants are examined. The paper discusses technical experience, identifies potential sources of technology for advanced concepts, cites the results of economic studies of mobile nuclear powerplants, and surveys future technical capabilities needed by examining the current use and projected needs for vehicles, machines, and habitats that could effectively use mobile nuclear reactor powerplants.

High Efficiency Heat Exchanger for High Temperature and High Pressure Applications

DOE Office of Scientific and Technical Information (OSTI.GOV)

Sienicki, James J.; Lv, Qiuping; Moisseytsev, Anton

CompRex, LLC (CompRex) specializes in the design and manufacture of compact heat exchangers and heat exchange reactors for high temperature and high pressure applications. CompRex’s proprietary compact technology not only increases heat exchange efficiency by at least 25 % but also reduces footprint by at least a factor of ten compared to traditional shell-and-tube solutions of the same capacity and by 15 to 20 % compared to other currently available Printed Circuit Heat Exchanger (PCHE) solutions. As a result, CompRex’s solution is especially suitable for Brayton cycle supercritical carbon dioxide (sCO2) systems given its high efficiency and significantly lower capitalmore » and operating expenses. CompRex has already successfully demonstrated its technology and ability to deliver with a pilot-scale compact heat exchanger that was under contract by the Naval Nuclear Laboratory for sCO2 power cycle development. The performance tested unit met or exceeded the thermal and hydraulic specifications with measured heat transfer between 95 to 98 % of maximum heat transfer and temperature and pressure drop values all consistent with the modeled values. CompRex’s vision is to commercialize its compact technology and become the leading provider for compact heat exchangers and heat exchange reactors for various applications including Brayton cycle sCO2 systems. One of the limitations of the sCO2 Brayton power cycle is the design and manufacturing of efficient heat exchangers at extreme operating conditions. Current diffusion-bonded heat exchangers have limitations on the channel size through which the fluid travels, resulting in excessive solid material per heat exchanger volume. CompRex’s design allows for more open area and shorter fluid proximity for increased heat transfer efficiency while sustaining the structural integrity needed for the application. CompRex is developing a novel improvement to its current heat exchanger design where fluids are directed to alternating channels so that each fluid is fully surrounded by the opposing fluid. As compared to similar existing compact heat exchangers, the new design converts most secondary surface area to primary surface area, eliminating fin inefficiencies. CompRex requests that all technical information about the heat exchanger designs be protected as proprietary information. To honor that request, only non-proprietay summaries are included in this report.« less

Status of the tokamak program

NASA Astrophysics Data System (ADS)

Sheffield, J.

1981-08-01

For a specific configuration of magnetic field and plasma to be economically attractive as a commercial source of energy, it must contain a high-pressure plasma in a stable fashion while thermally isolating the plasma from the walls of the containment vessel. The tokamak magnetic configuration is presently the most successful in terms of reaching the considered goals. Tokamaks were developed in the USSR in a program initiated in the mid-1950s. By the early 1970s tokamaks were operating not only in the USSR but also in the U.S., Australia, Europe, and Japan. The advanced state of the tokamak program is indicated by the fact that it is used as a testbed for generic fusion development - for auxiliary heating, diagnostics, materials - as well as for specific tokamak advancement. This has occurred because it is the most economic source of a large, reproducible, hot, dense plasma. The basic tokamak is considered along with tokamak improvements, impurity control, additional heating, particle and power balance in a tokamak, aspects of microscopic transport, and macroscopic stability.

Experimental study of heating scheme effect on the inner divertor power footprint widths in EAST lower single null discharges

NASA Astrophysics Data System (ADS)

Deng, G. Z.; Xu, J. C.; Liu, X.; Liu, X. J.; Liu, J. B.; Zhang, H.; Liu, S. C.; Chen, L.; Yan, N.; Feng, W.; Liu, H.; Xia, T. Y.; Zhang, B.; Shao, L. M.; Ming, T. F.; Xu, G. S.; Guo, H. Y.; Xu, X. Q.; Gao, X.; Wang, L.

2018-04-01

A comprehensive work of the effects of plasma current and heating schemes on divertor power footprint widths is carried out in the experimental advanced superconducting tokamak (EAST). The divertor power footprint widths, i.e., the scrape-off layer heat flux decay length λ q and the heat spreading S, are crucial physical and engineering parameters for fusion reactors. Strong inverse scaling of λ q and S with plasma current have been demonstrated for both neutral beam (NB) and lower hybrid wave (LHW) heated L-mode and H-mode plasmas at the inner divertor target. For plasmas heated by the combination of the two kinds of auxiliary heating schemes (NB and LHW), the divertor power widths tend to be larger in plasmas with higher ratio of LHW power. Comparison between experimental heat flux profiles at outer mid-plane (OMP) and divertor target for NB heated and LHW heated L-mode plasmas reveals that the magnetic topology changes induced by LHW may be the main reason to the wider divertor power widths in LHW heated discharges. The effect of heating schemes on divertor peak heat flux has also been investigated, and it is found that LHW heated discharges tend to have a lower divertor peak heat flux compared with NB heated discharges under similar input power. All these findings seem to suggest that plasmas with LHW auxiliary heating scheme are better heat exhaust scenarios for fusion reactors and should be the priorities for the design of next-step fusion reactors like China Fusion Engineering Test Reactor.

PREPARATION OF HIGH-DENSITY THORIUM OXIDE SPHERES

DOEpatents

McNees, R.A. Jr.; Taylor, A.J.

1963-12-31

A method of preparing high-density thorium oxide spheres for use in pellet beds in nuclear reactors is presented. Sinterable thorium oxide is first converted to free-flowing granules by means such as compression into a compact and comminution of the compact. The granules are then compressed into cubes having a density of 5.0 to 5.3 grams per cubic centimeter. The cubes are tumbled to form spheres by attrition, and the spheres are then fired at 1250 to 1350 deg C. The fired spheres are then polished and fired at a temperature above 1650 deg C to obtain high density. Spherical pellets produced by this method are highly resistant to mechanical attrition hy water. (AEC)

Current drive by spheromak injection into a tokamak

DOE Office of Scientific and Technical Information (OSTI.GOV)

Brown, M.R.; Bellan, P.M.

1990-04-30

We report the first observation of current drive by injection of a spheromak plasma into a tokamak (Caltech ENCORE small reasearch tokamak) due to the process of helicity injection. After an abrupt 30% increase, the tokamak current decays by a factor of 3 due to plasma cooling caused by the merging of the relatively cold spheromak with the tokamak. The tokamak density profile peaks sharply due to the injected spheromak plasma ({ital {bar n}}{sub 3} increases by a factor of 6) then becomes hollow, suggestive of an interchange instability.

First AC loss test and analysis of a Bi2212 cable-in-conduit conductor for fusion application

NASA Astrophysics Data System (ADS)

Qin, Jinggang; Shi, Yi; Wu, Yu; Li, Jiangang; Wang, Qiuliang; He, Yuxiang; Dai, Chao; Liu, Fang; Liu, Huajun; Mao, Zhehua; Nijhuis, Arend; Zhou, Chao; Devred, Arnaud

2018-01-01

The main goal of the Chinese fusion engineering test reactor (CFETR) is to build a fusion engineering tokamak reactor with a fusion power of 50-200 MW, and plan to test the breeding tritium during the fusion reaction. This may require a maximum magnetic field of the central solenoid and toroidal field coils up to 15 T. New magnet technologies should be developed for the next generation of fusion reactors with higher requirements. Bi2Sr2CaCu2Ox (Bi2212) is considered as a potential and promising superconductor for the magnets in the CFETR. R&D activities are ongoing at the Institute of Plasma Physics, Chinese Academy of Sciences for demonstration of the feasibility of a CICC based on Bi2212 round wire. One sub-size conductor cabled with 42 wires was designed, manufactured and tested with limited strand indentation during cabling and good transport performance. In this paper, the first test results and analysis on the AC loss of Bi2212 round wires and cabled conductor samples are presented. Furthermore, the impact of mechanical load on the AC loss of the sub-size conductor is investigated to represent the operation conditions with electromagnetic loads. The first tests provide an essential basis for the validation of Bi2212 CICC and its application in fusion magnets.

Reactor plasma facing component designs based on liquid metal concepts supported in porous systems

NASA Astrophysics Data System (ADS)

Tabarés, F. L.; Oyarzabal, E.; Martin-Rojo, A. B.; Tafalla, D.; de Castro, A.; Soleto, A.

2017-01-01

The use of liquid metals (LMs) as plasma facing components in fusion devices was proposed as early as 1970 for a field reversed concept and inertial fusion reactors. The idea was extensively developed during the APEX Project, at the turn of the century, and it is the subject at present of the biennial International Symposium on Lithium Applications (ISLA), whose fourth meeting took place in Granada, Spain at the end of September 2015. While liquid metal flowing concepts were specially addressed in USA research projects, the idea of embedding the metal in a capillary porous system (CPS) was put forwards by Russian teams in the 1990s, thus opening the possibility of static concepts. Since then, many ideas and accompanying experimental tests in fusion devices and laboratories have been produced, involving a large fraction of countries within the international fusion community. Within the EUROFusion Roadmap, these activities are encompassed into the working programs of the plasma facing components (PFC) and divertor tokamak test (DTT) packages. In this paper, a review of the state of the art in concepts based on the CPS set-up for a fusion reactor divertor target, aimed at preventing the ejection of the liquid metal by electro-magnetic (EM) forces generated under plasma operation, is described and required R+D activities on the topic, including ongoing work at CIEMAT specifically oriented to filling the remaining gaps, are stressed.

Individual dose due to radioactivity accidental release from fusion reactor.

PubMed

Nie, Baojie; Ni, Muyi; Wei, Shiping

2017-04-05

As an important index shaping the design of fusion safety system, evaluation of public radiation consequences have risen as a hot topic on the way to develop fusion energy. In this work, the comprehensive public early dose was evaluated due to unit gram tritium (HT/HTO), activated dust, activated corrosion products (ACPs) and activated gases accidental release from ITER like fusion reactor. Meanwhile, considering that we cannot completely eliminate the occurrence likelihood of multi-failure of vacuum vessel and tokamak building, we conservatively evaluated the public radiation consequences and environment restoration after the worst hypothetical accident preliminarily. The comparison results show early dose of different unit radioactivity release under different conditions. After further performing the radiation consequences, we find it possible that the hypothetical accident for ITER like fusion reactor would result in a level 6 accident according to INES, not appear level 7 like Chernobyl or Fukushima accidents. And from the point of environment restoration, we need at least 69 years for case 1 (1kg HTO and 1000kg dust release) and 34-52years for case 2 (1kg HTO and 10kg-100kg dust release) to wait the contaminated zone drop below the general public safety limit (1mSv per year) before it is suitable for human habitation. Copyright © 2016 Elsevier B.V. All rights reserved.

U. S. fusion programs: Struggling to stay in the game

DOE Office of Scientific and Technical Information (OSTI.GOV)

Crawford, M.

Funding for the US fusion energy program has suffered and will probably continue to suffer major cuts. A committee hand-picked by Energy Secretary James Watkins urged the Department of Energy to mount an aggressive program to develop fusion power, but congress cut funding from $323 million in 1990 to $275 million in 1991. This portends dire conditions for fusion research and development. Projects to receive top priority are concerned with the tokamaks and to keep the next big machine, the Burning Plasma Experiment, scheduled for beginning of construction in 1993 on schedule. Secretary Watkins is said to want to keepmore » the International Thermonuclear Energy Reactor (ITER) on schedule. ITER would follow the Burning Plasma Experiment.« less

Ethanol dehydration to ethylene in a stratified autothermal millisecond reactor.

PubMed

Skinner, Michael J; Michor, Edward L; Fan, Wei; Tsapatsis, Michael; Bhan, Aditya; Schmidt, Lanny D

2011-08-22

The concurrent decomposition and deoxygenation of ethanol was accomplished in a stratified reactor with 50-80 ms contact times. The stratified reactor comprised an upstream oxidation zone that contained Pt-coated Al(2)O(3) beads and a downstream dehydration zone consisting of H-ZSM-5 zeolite films deposited on Al(2)O(3) monoliths. Ethanol conversion, product selectivity, and reactor temperature profiles were measured for a range of fuel:oxygen ratios for two autothermal reactor configurations using two different sacrificial fuel mixtures: a parallel hydrogen-ethanol feed system and a series methane-ethanol feed system. Increasing the amount of oxygen relative to the fuel resulted in a monotonic increase in ethanol conversion in both reaction zones. The majority of the converted carbon was in the form of ethylene, where the ethanol carbon-carbon bonds stayed intact while the oxygen was removed. Over 90% yield of ethylene was achieved by using methane as a sacrificial fuel. These results demonstrate that noble metals can be successfully paired with zeolites to create a stratified autothermal reactor capable of removing oxygen from biomass model compounds in a compact, continuous flow system that can be configured to have multiple feed inputs, depending on process restrictions. Copyright © 2011 WILEY-VCH Verlag GmbH & Co. KGaA, Weinheim.