Monte Carlo Neutral Transport Research Articles

Progress towards the validation of models of the behavior of neutral helium in gas puff imaging experiments

Gas puff imaging (GPI) experiments are designed to provide high time and space resolution data on the structure of plasma turbulence in the plane perpendicular to the magnetic field. We first examine the temporal behavior of the helium atoms used as the emitting species for GPI and show that for the time scales of interest (≳1μs), the atomic physics model underlying the conventional interpretation of GPI is valid. Second, we continue the Monte Carlo neutral transport simulations of the GPI diagnostic begun in [D.P. Stotler et al., Contrib. Plasma Phys. 44 (2004) 294]. The radial characteristics of the simulated emission clouds match observations to within the estimated errors. The upshot of these two results is that the technique for unfolding the 2-D, time-dependent plasma density and temperature data from helium GPI emission, relying on this atomic physics model and utilizing the simulated neutral density data from DEGAS 2, is valid.

Towards a revised Monte Carlo neutral particle surface interaction model

The components of the neutral- and plasma-surface interaction model used in the Monte Carlo neutral transport code DEGAS 2 are reviewed. The idealized surfaces and processes handled by that model are inadequate for accurately simulating neutral transport behaviour in present day and future fusion devices. We identify some of the physical processes missing from the model, such as mixed materials and implanted hydrogen, and make some suggestions for improving the model.

Three-dimensional simulation of gas conductance measurement experiments on Alcator C-Mod

Three-dimensional Monte Carlo neutral transport simulations of gas flow through the Alcator C-Mod sub-divertor yield conductances comparable to those found in dedicated experiments. All are significantly smaller than the conductance found with the previously used axisymmetric geometry. A benchmarking exercise of the code against known conductance values for gas flow through a simple pipe provides a physical basis for interpreting the comparison of the three-dimensional and experimental C-Mod conductances.

Shield structure optimisation studies for the west beam port of the KAMINI reactor

KAMINI is the Kalpakkam Mini Reactor and its main purpose is to cater to the experimental needs and for neutron radiography. It is a water-cooled reactor with 233 U as the fissile material. The reactor has three neutron beam ports for experimental needs, which are made of graded cylindrical aluminium channel. All the beam ports start from the core, pierce the biological shield and are 2 m long. The shield structure optimisation studies for the beam port towards the west side of the reactor are presented here. The diameter of the west beam channel at the core centre is 54 mm and at the other end is 250 mm. The west beam tube opening is 530 mm below the floor level and hence the pit housing the experimental cavity is below the floor level with dimensions 2 m × 2.5 m × 1.3 m. The beam tube opening into the experimental cavity serves as the neutron source for radiation physics experiments and is assumed as a surface source in the calculations. Rough estimate of the shield design is made based on the literature on dose-equivalent index transmission through concrete for average neutron energy of 1.5 MeV. Detailed radiation transport calculations are performed using Monte Carlo neutral particle transport code (MCNP) to optimise the shield design. Neutron and capture gamma dose rates at the accessible areas are estimated. The contribution of prompt fission gamma rays is found to be negligible compared to the dose rates due to capture gamma rays. The details of the optimised shield structure proposed for the west beam port are as follows. Fixed concrete shields of thickness 650 mm on the lateral sides and a composite shield (500 mm paraffin and 50 mm concrete) in the front side at a distance of 1 m from the beam tube opening are recommended inside the experimental pit. During reactor operation, a composite mobile shield (500 mm paraffin and 500 mm concrete) closes the experimental cavity at the floor level. Fixed concrete shields are recommended to close the pit fully. The shield structure proposed increases the experimental cavity volume from 0.2 to 1.4 m 3 with the dose levels at the accessible areas less than one μSv/h. The MCNP computed neutron and gamma dose rates are compared with the measured values with the existing shield structure to verify the source term used.

Numerical modeling of excited hydrogen atoms and their transport in hydrogen negative ion sources

Improvements are made of a Monte Carlo neutral transport code for the analysis of negative ion sources. These improvements make it possible to calculate the density of excited atoms, and to compare the numerical results with spectroscopic measurements, such as the Hα Balmer spectrum (n=3→n=2, n: principle number). The line-integrated intensities of the Hα spectrum in a tandem-type negative ion source are investigated. The excited hydrogen atoms in the n=3 level are populated by excitation, mutual neutralization (H−+H+→H(n)+H), and dissociation of H2, etc. The simulation results show that mutual neutralization greatly contributes to the intensity of the Hα spectrum, especially in the extraction region. In addition, near the filament in the driver region, the excitation process due to the fast electrons greatly contributes to the intensity of the Hα spectrum.

Shield structure optimisation studies for the west beam port of the KAMINI reactor

KAMINI is the Kalpakkam Mini Reactor and its main purpose is to cater to the experimental needs and for neutron radiography. It is a water-cooled reactor with 233U as the fissile material. The reactor has three neutron beam ports for experimental needs, which are made of graded cylindrical aluminium channel. All the beam ports start from the core, pierce the biological shield and are 2 m long. The shield structure optimisation studies for the beam port towards the west side of the reactor are presented here. The diameter of the west beam channel at the core centre is 54 mm and at the other end is 250 mm. The west beam tube opening is 530 mm below the floor level and hence the pit housing the experimental cavity is below the floor level with dimensions 2 m × 2.5 m × 1.3 m. The beam tube opening into the experimental cavity serves as the neutron source for radiation physics experiments and is assumed as a surface source in the calculations. Rough estimate of the shield design is made based on the literature on dose-equivalent index transmission through concrete for average neutron energy of 1.5 MeV. Detailed radiation transport calculations are performed using Monte Carlo neutral particle transport code (MCNP) to optimise the shield design. Neutron and capture gamma dose rates at the accessible areas are estimated. The contribution of prompt fission gamma rays is found to be negligible compared to the dose rates due to capture gamma rays. The details of the optimised shield structure proposed for the west beam port are as follows. Fixed concrete shields of thickness 650 mm on the lateral sides and a composite shield (500 mm paraffin and 50 mm concrete) in the front side at a distance of 1 m from the beam tube opening are recommended inside the experimental pit. During reactor operation, a composite mobile shield (500 mm paraffin and 500 mm concrete) closes the experimental cavity at the floor level. Fixed concrete shields are recommended to close the pit fully. The shield structure proposed increases the experimental cavity volume from 0.2 to 1.4 m 3 with the dose levels at the accessible areas less than one μSv/h. The MCNP computed neutron and gamma dose rates are compared with the measured values with the existing shield structure to verify the source term used.

Three‐Dimensional Neutral Transport Simulations of Gas Puff Imaging Experiments

AbstractGas Puff Imaging (GPI) experiments are designed to isolate the structure of plasma turbulence in the plane perpendicular to the magnetic field. Three‐dimensional aspects of this diagnostic technique as used on the National Spherical Torus eXperiment (NSTX) are examined via Monte Carlo neutral transport simulations. The radial width of the simulated GPI images are in rough agreement with observations. However, the simulated emission clouds are angled approximately 15° with respect to the experimental images. The simulations indicate that the finite extent of the gas puff along the viewing direction does not significantly degrade the radial resolution of the diagnostic. These simulations also yield effective neutral density data that can be used in an approximate attempt to infer 2‐D electron density and temperature profiles from the experimental images. (© 2004 WILEY‐VCH Verlag GmbH & Co. KGaA, Weinheim)

Neutral transport simulations of gas puff imaging experiments

Visible imaging of gas puffs has been used on the Alcator C-Mod tokamak to characterize edge plasma turbulence, yielding data that can be compared with plasma turbulence codes. Simulations of these experiments with the DEGAS 2 Monte Carlo neutral transport code have been carried out to explore the relationship between the plasma fluctuations and the observed light emission. By imposing two-dimensional modulations on the measured time-average plasma density and temperature profiles, we demonstrate that the spatial structure of the emission cloud reflects that of the underlying turbulence. However, the photon emission rate depends on the plasma density and temperature in a complicated way, and no simple scheme for inferring the plasma parameters directly from the light emission patterns is apparent. The simulations indicate that excited atoms generated by molecular dissociation are a significant source of photons, further complicating interpretation of the gas puff imaging results.

Neutral transport baffling studies for SST-1

The Superconducting Steady-State Tokamak (SST-1) is being designed to have 1000 s shots. Scrape off layer (SOL) modelling of SST-1 shows a sheath limited flow regime in the SOL with low recycling of neutrals in the divertor chamber. This implies that there will be a high neutral backflow from the divertor plates to the core which can result in decreased density control or plasma disruption unless the neutrals recycling from the divertor plates are well baffled to prevent them from entering the core. This paper describes the various baffle plate designs considered for SST-1 and the procedure by which a suitable design was accepted. The Monte Carlo neutral transport code DEGAS was used to study the neutral transport and a steady state particle balance model was devised to interpret the DEGAS results. It is seen that fueling efficiency is an important parameter which should be considered when calculating the throughput at the pump opening and the particle backflow to the core. It is also seen that high recycling alone does not ensure a good density control, it is the ratio of the particle throughput at the pumps to the particle backflow which has to be optimised for good density control. The SST-1 baffle plates are designed to minimise the particle backflow to the core to achieve good density control, keeping the pumping requirements within bounds.

Modeling of Alcator C-Mod divertor baffling experiments

A specific Alcator C-Mod discharge from the series of divertor baffling experiments is simulated with the DEGAS 2 Monte Carlo neutral transport code. A simple two-point plasma model is used to describe the plasma variation between Langmuir probe locations. A range of conductances for the bypass between the divertor plenum and the main chamber are considered. The experimentally observed insensitivity of the neutral current flowing through the bypass and of the D α emissions to the magnitude of the conductance is reproduced. The current of atoms in this regime is being limited by atomic physics processes and not bypass conductance. The simulated trends in divertor pressure, bypass current, and D α emission agree only qualitatively with the experimental measurements, however. Possible explanations for the quantitative differences are discussed.

Improved elastic collision modeling in DEGAS 2 for low-temperature plasmas

Recent emphasis on low-temperature divertor operations has focused attention on proper treatment of neutral-elastic collisions in low-temperature environments. For like species collisions, as in D++D, quantum mechanical indistinguishability precludes differentiation of small-angle elastic scattering from resonant charge exchange for collision energies <2 eV. The current work improves the low-temperature simulation capabilities of the DEGAS 2 Monte Carlo neutral transport code [D. P. Stotler and C. F. F. Karney, Contrib. Plasma Phys. 34, 392–397 (1994)] by employing a set of quantum mechanical collision cross sections produced by Oak Ridge National Laboratory to provide a comprehensive treatment of ion-neutral elastic collisions. Ion-molecular collisions in the form of D++D2 are included for the first time. An integration technique is utilized that reduces the total collision cross section while keeping the other transport cross sections invariant. The inclusion of ion-molecular elastic collisions results in significant increases in energy exchange between background ions and neutral test species.

Formulas Giving Buildup Factor for Double-Layered Shields

Formulas that give absorbed dose buildup factors for two-layered shields have been developed based on gamma-ray absorption buildup factors computed with the Monte Carlo Neutral Particle Transport Code System (MCNP). The shielding materials considered were water, lead, steel, concrete, and some of their combinations for two-layered shields with thicknesses between 1 to 10 mfp. Gamma energy considered ranged from 0.5 to 6 MeV. The formulas reproduce MCNP results with a difference of <10%, in most cases <3%.

Design of the KSTAR Divertor

The planned Korea Superconducting Tokamak Advanced Research (KSTAR) divertor has been designed to provide reliable power handling and particle control with enough shaping flexibility to accommodate a wide range of plasma operation. The physics basis for the current configuration of the KSTAR divertor through analyses of the heat flux at the target, particle control, and plasma-facing component is reported. A simple zero-dimensional model based on the power balance assumptions and two-dimensional codes is utilized to estimate the heat flux to the divertor plate. The limit for the peak heat flux on the divertor plate, 3.5 MW/m2, requires advanced operating modes such as the radiative divertor and radiative mantle, which are considered to overcome the weakness of a high-recycling divertor. A simple particle balance model could estimate the pumping rate with total leakage fraction assuming particle sources. A Monte Carlo neutral transport calculation determines the dimension of a gap between the center and outer divertor targets. It also determines the number and best position of the pumps, as well as the geometry for conductance. For the initial 20-s discharges, a bolted-tile carbon-fiber-composite design is relied upon for the upper and lower divertor targets. The design of the supporting structure for the divertors will allow for future modifications to accommodate thermal steady-state 300-s operation or to optimize divertor performance based on new understanding gained during initial tokamak operation.

B2-EIRENE code modelling of an island divertor

The multifluid plasma scrape-off layer transport code B2 coupled to the EIRENE Monte Carlo neutral gas transport code was applied to a prototypical island divertor configuration in W7-AS. The approach predicts the accessibility of stable, layer-like energy detachment within a relatively broad density range. The island geometry with typically small field line pitch has the beneficial effects of efficient shielding of the main plasma against neutrals, and of strong momentum losses from the power carrying layer due to momentum radial transport. The latter reduces the peak power load and eases detachment. The power exhaust properties and the onset of detachment can be positively affected by the choice of the field line pitch which can be varied in W7-AS by superimposed control fields.

Coupled Monte Carlo neutral – fluid plasma simulation of Alcator C-Mod divertor plasma near detachment

Using the coupled fluid plasma and Monte Carlo neutral transport code, B2-EIRENE, we simulate and analyze a nearly detached Alcator C-Mod discharge exhibiting the divertor over-pressure or “death ray” phenomenon. Qualitative agreement is obtained between the measured and simulated density and temperature profiles at the outboard midplane and divertor target. In the simulation, the divertor over-pressure is caused by radial transport of parallel momentum. Momentum transport by the neutral species is negligible.

Scrape off layer modelling studies for SST-I

SOL modelling results for SST-1 (SST Team, Proceedings of the 16th IEEE/NPSS Symposium on Fusion Engineering, Champaign, IL, vol. II, 1995, p. 481) show a sheath limited flow regime. This is due to the low edge densities required by lower hybrid current drive (LHCD), coupled with high power input per unit volume. Coupled plasma–neutral transport studies using B2-Eirene [R. Schneider et al., J. Nucl. Mater. 196–198 (1992) 810] show significantly high charge exchange losses and radiated power from the core. It also shows that the heat flux to the inner divertor is higher than that to the outer divertor due to thinner inner SOL widths. The Monte-Carlo neutral transport code DEGAS [D. Heifitz et al., J. Comput. Phys. 46 (1982) 309] was used to optimise the baffle plate geometry and it was seen that a configuration where the baffle plate shields the main plasma from the divertor strike point results in reduced backflow of neutrals. The divertor erosion code DIVER (M. Warrier et al., SST Divertor Modelling Report, 1996–1997) was used to predict a steady state operating temperature for the SST divertor plate lying in the range 750–1000°C for which the erosion will be minimum.

Modeling of neutral plasma in a divertor in the fluid‐kinetic transition

AbstractThe neutrals in a tokamak divertor in the so‐called “detached” mode of operation may be in a kinetic regime (long mean free path) in some regions and in a fluid regime (short mean free path) elsewhere. For example, a kinetic description is necessary in the thin ionization front, while the cold neutrals in front of the divertor plate are adequately described by fluid equations. The paper describes a way to model this by coupling existing Monte Carlo neutral transport and plasma fluid codes. This is explored with a simple analytical model.

Study of helium transport in HL-1M edge plasma region

Radial profiles of the emission intensity for H α and neutral helium line ( λ = 5875.7 A ̊ ) are measured to study helium transport and recycling in the HL-1M edge region in the vicinity of the poloidal graphite limiter (at r = 26 cm) after previous helium glow discharge cleaning and boronization under different LHCD and ECH conditions. A one-dimensional Monte Carlo neutral helium transport code, 1DHET, is utilized to calculate neutral helium density profile in the same region. The comparisons show that the freshly boronized wall with helium glow discharge cleaning exhibits somewhat selective recycling characteristics between the helium and hydrogen atoms, if ECH power is not applied.

Modeling of neutral hydrogen velocities in the Tokamak Fusion Test Reactor

Monte Carlo neutral transport simulations of hydrogen velocities in the Tokamak Fusion Test Reactor (TFTR) [K. M. McGuire et al., Phys. Plasmas 2, 2176 (1995)] are compared with experiment using the Doppler-broadened Balmer-α spectral line profile. Good agreement is obtained under a range of conditions, validating the treatment of charge exchange, molecular dissociation, surface reflection, and sputtering in the neutral gas code DEGAS [D. Heifetz et al., J. Comput. Phys. 46, 309 (1982)]. A residual deficiency of 10–100 eV neutrals in most of the simulations indicates that further study of the energetics of H+2 dissociation for electron energies in excess of 100 eV is needed.

Sensitivity of Neutral Throughput to Geometry and Plasma Position in the Tokamak Physics Experiment Divertor Region

Sufficient neutral atom and molecular throughput is essential for the steady-state operation of the proposed Tokamak Physics Experiment tokamak. To predict the throughput, the B2 edge-plasma fluid code and the DEGAS Monte Carlo neutral transport code were coupled globally. For the day 1 low-power (17.5-MW) operation condition, the recycling coefficient for both codes matched at 0.985, implying that for every 1000 ions striking the divertor plate, 15 are ultimately removed down the pump duct. The neutral molecular density was 2.52{+-}0.15x10{sup 19}m{sup 3}, giving a throughput of 92.6{+-}5.6 Torr.l/s. Varying the scrape-off length for the plasma extending into the gap between the baffle and the plate from 0 to 2 cm decreased the throughput by a factor of >2. Moving the strike point away from the gap at first increases the throughput by lessening the pumping efficiency of the plasma in the gap. As the plasma is moved even farther away, the throughput drops due to a lack of source term for neutrals entering the pumped region. Illustrating the importance of moving the source term, moving the strike point away from the gap but retaining the original plasma in the gap lowers the throughput by a factor of 10. Altering themore » curvature of the baffle has little effect on the neutral solution. 7 refs., 9 figs.« less