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Erosion/redeposition analysis of the ITER (International Tokamak Engineering Reactor) divertor

Description: Sputtering erosion of the proposed ITER divertor has been analyzed using the REDEP computer code. A carbon coated plate at medium and low plasma edge temperatures, as well as beryllium and tungsten plates, have been examined. Peak net erosion rates for C and Be are very high (/approximately/20--80 cm/burn/center dot/yr) though an order of magnitude less than the gross rates. Tritium buildup rates in codeposited carbon surface layers may also be high (/approximately/50--250 kg/burn/center dot/yr). Plasma contamination, however, from divertor sputtering is low (/approx lt/.5%). Operation with low Z divertor plates, at high duty factors, therefore appears unacceptable due to erosion, but may work for low duty factor (/approximately/2%) ''physics phase'' operation. Sweeping of the poloidal field lines at the divertor can reduce erosion, by typical factors of /approximately/2--8. A tungsten coated plate works well, from the erosion standpoint, for plasma plate temperatures of /approximately/40 eV or less. 18 refs., 11 figs., 3 tabs.
Date: July 1, 1989
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Temperature limit of a graphite divertor surface due to particle erosion

Description: The surface temperature limit of a graphite coated tokamak divertor due to particle erosion has been examined theoretically. The analysis uses models for physical sputtering, radiation enhanced sublimation, and redeposition phenomena. An upper bound on surface temperature is computed based on the necessary condition that self-sputtering be finite. The allowable graphite surface temperature is found to depend strongly on the plasma edge temperature, being limited to /approx lt/1150 C for medium plasma edge temperatures (/approximately/150 eV). 13 refs., 2 figs., 1 tab.
Date: March 1, 1989
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Kinetic solution of the sheath region in a fusion reactor

Description: The sheath region in a fusion reactor is studied with a one-dimensional kinetic code. The sheath potential, heat transmission, and sputtering of the boundary are all quite sensitive to electron re-emission. The expected heat and particle fluxes in future fusion reactors leads to a prediction of keV-edge temperatures.
Date: December 1, 1979
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Self-consistent calculations of edge temperature and self-sputtering of the limiter surface for tokamak fusion reactors

Description: Self-sputtering and edge temperature estimates have been made for a tokamak fusion reactor with a normal incidence boundary using models for the power balance, plasma sheath, charge state, and sputtering coefficients. Both structural materials and low-Z coatings have been examined. From the self-sputtering standpoint, some materials will work only at very low edge temperatures; these will require a high recycling rate and a high radiation fraction.
Date: May 1, 1980
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

New impurity control methods for tokamak fusion power reactors

Description: Two methods of impurity control, a helium retention wall and a limiter/wall/vacuum system, are analyzed as candidates for experimental and commercial tokamak fusion reactors. Models were developed for both of these systems and used to compute the steady-state sputtered impurity and helium concentrations as well as the burn-cycle power performance obtainable with these systems. It has been found that relatively modest requirements are needed, for both of these systems, to obtain a non-impurity limited burn. In addition, power output approaching the beta-limited maximum can be obtained with the limiter system if enough of the particle flux for the limiter can be directed away from the plasma.
Date: January 1, 1978
Creator: Brooks, J N
Partner: UNT Libraries Government Documents Department

Sputtering performance of the TFCX limiter

Description: The sputtering performance of the proposed TFCX pumped limiter was analyzed using the REDEP computer code. Erosion, redeposition, surface shape and heat flux changes with time, and plasma contamination issues were examined. A carbon coated limiter was found to give acceptable sputtering performance over the TFCX lifetime if, and only if, acceptable redeposition properties of carbon are obtained.
Date: September 1, 1984
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Plasma sheath region near a boundary with positive ion backscattering

Description: The sheath region at a boundary perpendicular to a magnetic field has been studied by numerically solving time-independent Vlasov-Poisson equations for a DT plasma. Boundary conditions are specified for particles entering the sheath region and for particles returning from the boundary. The effect of hydrogen ion backscatter as well as electron reemission has been examined. For a space-charge limited electron reemission condition, even a small amount of ion backscatter can change the sheath parameters substantially, lowering the sheath potential and increasing the heat transmission. This could be important in lowering the edge temperature, and the sputtering rate in future fusion devices.
Date: April 1, 1981
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Energy deposition in STARFIRE reactor components

Description: The energy deposition in the STARFIRE commercial tokamak reactor was calculated based on detailed models for the different reactor components. The heat deposition and the 14 MeV neutron flux poloidal distributions in the first wall were obtained. The poloidal surface heat load distribution in the first wall was calculated from the plasma radiation. The Monte Carlo method was used for the calculation to allow an accurate modeling for the reactor geometry.
Date: April 1, 1985
Creator: Gohar, Y. & Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Transport calculations of chemically sputtered carbon near a plasma divertor surface

Description: The transport of chemically sputtered carbon near a tokamak divertor surface has been analyzed with the Monte Carlo code WBC. The code follows the motion of sputtered methane atoms and the resulting carbon and hydrocarbon derivatives. Ion transport due to the magnetic field, sheath electric field, and collisions with the plasma is computed. Redeposition fractions, impinging species type, charge state, and velocity have been analyzed. For plasma temperatures {ge} 10 eV, and for typical divertor plasma densities, local redeposition of chemically sputtered carbon approaches 100%. Redeposition fractions are lower ({approximately}80%) for lower temperatures and/or lower density. Physical sputtering of carbon due to redeposition of chemically sputtered material is low but a hydrocarbon reflection cascade due to redeposition may be high.
Date: January 1, 1992
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Divertor erosion study for TPX and implications for steady-state fusion reactors

Description: A sputtering erosion analysis was performed for the tilted plate divertor design of the proposed TPX tokamak. High temperature ({approximately} 100 eV), non-radiative, steady-state compatible, plasma edge conditions were used as input to the REDEP erosion/redeposition code. For the reference carbon surface the results show a stable erosion profile, i.e., non-runaway self-sputtering, in spite of carbon self-sputtering coefficients that are locally in excess of unity. The resulting net erosion rates are high (peak {approx} 1--2.5 m/burn-yr) but may be acceptable for a low duty factor experimental device such as TPX. Other surface materials were also analyzed, in part to obtain insight for fusion reactor designs using a similar plasma regime. Both medium and high-Z materials are predicted not to work, due to runaway self-sputtering. Beryllium is stable but has erosion rates as high or higher than carbon. A liquid metal lithium surface has stable sputtering with a zero-erosion potential and may thus be an attractive future material choice.
Date: December 31, 1995
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Erosion/redeposition analysis : status of modeling and code validation for semi-detached tokamak edge plasmas.

Description: We are analyzing erosion and tritium codeposition for ITER, DIII-D, and other devices with a focus on carbon divertor and metallic wall sputtering, for detached and semi-detached edge plasmas. Carbon chemical-sputtering hydrocarbon-transport is computed in detail using upgraded models for sputtering yields, species, and atomic and molecular processes. For the DIII-D analysis this includes proton impact and dissociative recombination for the full methane and higher hydrocarbon chains. Several mixed material (Si-C doping and Be/C) effects on erosion are examined. A semi-detached reactor plasma regime yields peak net wall erosion rates of {approximately}1.0 (Be), {approximately}0.3 (Fe), and {approximately}0.01 (W) cm/burn-yr, and {approximately}50 cm/burn-yr for a carbon divertor. Net carbon erosion is dominated by chemical sputtering in the {approximately}1-3 eV detached plasma zone. Tritium codeposition in divertor-sputtered redeposited carbon is high ({approximately}10-20 g-T/1000 s ). Silicon and beryllium mixing tends to reduce carbon erosion. Initial hydrocarbon transport calculations for the DIII-D DiMES-73 detached plasma experiment show a broad spectrum of redeposited molecules with {approximately}90% redeposition fraction.
Date: January 19, 1999
Creator: Brooks, J. N.
Partner: UNT Libraries Government Documents Department

Redeposition of the sputtered surface in limiters

Description: Erosion of the surface coating of a pumped limiter by sputtering may be a critical life-limiting issue for future tokamak reactors. Redeposition of the sputtered material, however, may extend the coating life significantly. This subject has now been studied through the use of a code which models the redeposition of sputtered material which gets ionized in the scrape-off layer. The code also treats the transfer of wall-sputtered material to the limiter. The code uses models of the plasma density and temperature in the scrape-off zone, sheath potential, sputtering coefficients, spatial distribution of the sputtered atoms, and electron impact ionization coefficient for the sputtered atoms. The studies were made for high flux and low flux edge conditions corresponding to FED and STARFIRE limiters and assumed plasma-edge parameters. The results indicate that substantial redeposition from the scrape-off layer ionized neutrals occurs in the cases considered.
Date: January 1, 1981
Creator: Brooks, J.N. & McGrath, R.T.
Partner: UNT Libraries Government Documents Department

Low plasma edge temperatures for the self-pumped limiter

Description: Transport code calculations have been performed to study the operation of an INTOR-like tokamak plasma from which helium is removed by a self-pumped limiter, which traps helium, but not hydrogen, in its surface layers. To prevent saturation by helium, the surface is renewed by continuous injection of the surface material (vanadium in this study) into the scrape-off layer. The presence of the injected vanadium leads to plasma temperatures well below 50 eV in the scrape-off layer, with supplementary rf heating. Operation in this edge temperature regime is essential for the use of medium- and high-Z limiter coatings.
Date: March 1, 1985
Creator: Terry, W.K. & Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Burn cycle requirements comparison of pulsed and steady-state tokamak reactors

Description: Burn cycle parameters and energy transfer system requirements were analyzed for an 8-m commercial tokamak reactor using four types of cycles: conventional, hybrid, internal transformer, and steady state. Not surprisingly, steady state is the best burn mode if it can be achieved. The hybrid cycle is a promising alternative to the conventional. In contrast, the internal transformer cycle does not appear attractive for the size tokamak in question.
Date: December 1, 1983
Creator: Brooks, J.N. & Ehst, D.A.
Partner: UNT Libraries Government Documents Department

Self-pumping impurity control by in-situ metal deposition

Description: A system for in-situ removal of helium by trapping in freshly deposited metal surface layers of a limiter or divertor has been studied. The system would trap helium on a limiter front surface, or a divertor plate, at low plasma edge temperatures, or in a limiter slot region, at high edge temperatures. Fresh material, introduced to the plasma and/or scrape-off zone, would be added at a rate of about five times the alpha production rate. The material would be reprocessed periodically, e.g., once year. Possible materials are nickel, vanadium, niobium, and tantalum. Advantages of a self-pumping system are the absence of vacuum ducts and pumps, and the minimization of tritium processing and inventory.
Date: May 1, 1983
Creator: Brooks, J.N. & Mattas, R.F.
Partner: UNT Libraries Government Documents Department

Power supply requirements for a tokamak fusion reactor

Description: The power supply requirements for a 7-M major radius commercial tokamak reactor have been examined, using a system approach combining models of the reactor and poloidal coil set, plasma burn cycle and MHD calculations, and power supply characteristics and cost data. A conventional system using an MGF set and solid-state rectifier/inverter power supplies was studied in addition to systems using a homopolar generator, superconducting energy storage inductor, and dump resistors. The requirements and cost of the power supplies depend on several factors but most critically on the ohmic heating ramp time used for startup. Long ramp times (approx. > 8 s) seems to be feasible, from the standpoint of resistive volt-second losses, and would appear to make conventional systems quite competitive with nonconventional ones, which require further research and development.
Date: February 1, 1979
Creator: Brooks, J.N. & Kustom, R.L.
Partner: UNT Libraries Government Documents Department

Self-pumping impurity by in-situ metal deposition

Description: A system for in-situ removal of helium trapping in freshly deposited metal surface layers of a limiter or divertor has been studied. The system would trap helium on a limiter front surface, or a divertor plate, at low plasma edge temperatures, or in a limiter slot region, at high edge temperatures. Fresh material, introduced to the plasma and/or scrape-off zone, would be added at a rate of about five times the alpha production rate. The material would be reprocessed periodically, e.g. once a year. Possible materials are nickel, vanadium, niobium, and tantalum. Advantages of a self-pumping system are the absence of vacuum ducts and pumps, and the minimization of tritium processing and inventory.
Date: July 1, 1983
Creator: Brooks, J.N. & Mattas, R.F.
Partner: UNT Libraries Government Documents Department

ITER divertor sputtering erosion -- recent analysis for carbon, beryllium, tungsten, and niobium surfaces

Description: ITER divertor plate sputtering erosion has been analyzed using current design information and updated impurity transport models. The REDEP erosion/redeposition code was used to compute erosion for a very low plasma divertor temperature (T{sub e{sub 0}} {approx} 12 eV) physics phase reference'' case, and for other plasma conditions. A high surface temperature case (T{sub s{sub 0}} = 1800{degree}C) is analyzed for a carbon surface. Niobium is analyzed using WBC near-surface transport code results for the redeposited charge state. The REDEP results show high net erosion rates ({approx gt} 20 cm/burn{sm bullet}yr) for beryllium and carbon, even at low plasma temperatures. Net erosion rates are low to moderate for niobium ({approximately}0-3 cm/burn{sm bullet}yr), depending on plasma conditions, and low for tungsten ({approximately}0-0.2 cm/burn{sm bullet}yr). 9 refs., 2 figs.
Date: July 1, 1991
Creator: Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Study of plasma-edge conditions for a commercial tokamak reactor

Description: Particle and energy fluxes to the first wall and pumped limiter of a commercial-size tokamak power reactor have been studied with a one-dimensional, time-dependent plasma transport code. The plasma is operated in an enhanced radiation mode, with iodine as the high-Z impurity, whereby the transport power to the limiter is held to about half of the alpha-heating power. Recycling of neutral hydrogen and helium is followed in detail. The ion flux to the limiter and the neutral flux to the first wall are found to depend on the removal efficiency of the limiter/vacuum system. These fluxes also depend strongly on whether gas puffing or pellet injection is used for fueling.
Date: January 1, 1981
Creator: Boley, C.D. & Brooks, J.N.
Partner: UNT Libraries Government Documents Department

WILDCAT: a catalyzed D-D tokamak reactor

Description: WILDCAT is a conceptual design of a catalyzed D-D, tokamak, commercial, fusion reactor. WILDCAT utilizes the beneficial features of no tritium breeding, while not extrapolating unnecessarily from existing D-T designs. The reactor is larger and has higher magnetic fields and plasma pressures than typical D-T devices. It is more costly, but eliminates problems associated with tritium breeding and has tritium inventories and throughputs approximately two orders of magnitude less than typical D-T reactors. There are both a steady-state version with Alfven-wave current drive and a pulsed version. Extensive comparison with D-T devices has been made, and cost and safety analyses have been included. All of the major reactor systems have been worked out to a level of detail appropriate to a complete, conceptual design.
Date: November 1, 1981
Creator: Evans, K. Jr.; Baker, C.C. & Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Analysis of erosion and transport of carbon impurity in the TFTR inner bumper limiter region

Description: Carbon sputtering and transport on the TFTR inner graphite bumper limiter is investigated with the impurity transport code REDEP. Analysis is carried out for a series of ohmic discharges in TFTR. Predictions for Z{sub eff} in the core plasma agree well with in-situ experimental measurements. Run-away self-sputtering of carbon is predicted at low densities and high edge plasma temperatures when the limiter surface was purged of deuterium. Surface erosion and deposition is analyzed. In general, redeposition reduces the peak erosion by about a factor of five. Analysis is also carried out for a typical neutral beam heated discharge with a noncircular plasma. Spatial surface erosion and deposition profiles are compared qualitatively with beta backscattering measurements of metal deposition found on the limiter.
Date: January 1, 1992
Creator: Hua, T.Q. & Brooks, J.N.
Partner: UNT Libraries Government Documents Department

Impurity transport calculations for a drift-dependent tokamak scrape-off plasma

Description: Two dimensional calculations of impurtiy transport in a high recycling divertor scrape-off region have been made with an updated version of the ZTRANS Monte Carlo computer code. The calculations use plasma parameters for the Doublet 3 divertor, as computed by the Planet Fluid Transport Code. The effects of electric field, particle drift velocities, and thermal forces are included in the calculations. For all impurity species studied, it is found that impurity transport is dominated by frictional forces, over most of the scrape-off region. Light impurities, however, impinge substantially closer to the divertor plate center than do heavy impurities, which tend to impinge at the outer plate boundary. 8 refs., 4 figs., 2 tabs.
Date: April 1, 1988
Creator: Brooks, J.N.; Petravic, M. & Petravic, G.K.
Partner: UNT Libraries Government Documents Department

Plasma driving system requirements for commercial tokamak fusion reactors

Description: The plasma driving system for a tokamak reactor is composed of an ohmic heating (OH) coil, equilibrium field (EF) coil, and their respective power supplies. Conceptual designs of an Experimental Power Reactor (EPR) and scoping studies of a Demonstration Power Reactor have shown that the driving system constitutes a significant part of the overall reactor cost. The capabilities of the driving system also set or help set important parameters of the burn cycle, such as the startup time, and the net power output. Previous detailed studies on driving system dynamics have helped to define the required characteristics for fast-pulsed superconducting magnets, homopolar generators, and very high power (GVA) power supplies for an EPR. This paper summarizes results for a single reactor configuration together with several design concepts for the driving system. Both the reactor configuration and the driving system concepts are natural extensions from the EPR. Thus, the new results can be compared with the previous EPR results to obtain a consistent picture of how the driving system requirements will evolve--for one particular design configuration.
Date: January 1, 1977
Creator: Brooks, J.N.; Kustom, R.C. & Stacey, W.M. Jr.
Partner: UNT Libraries Government Documents Department