Monte Carlo Neutral Transport Code Research Articles

Development of Monte Carlo neutral transport code and the application to methane seeding plasma detachment

AbstractThe collisions between neutral particles and plasma in the divertor region determine divertor detachment and heat flux to the target. However, the diagnostic of neutral particle information in the tokamak experiment is very lacking. To study the behavior of neutral particles in the boundary region flexibly and comprehensively, a two‐dimensional kinetic neutral transport code based on Monte Carlo method is developed in this work. The code is applied to simulate the neutral particles in the boundary of EAST device. Firstly, the simulation results are compared with EIRENE using an identical plasma background, and good agreement is obtained. Secondly, the transport of methane and the corresponding hydrocarbon compounds is simulated using the developed code. The methane injection case is compared with the injection of carbon and deuterium molecule mixture gas (ratio 1:2) to investigate their impact on detachment and fueling. The simulation reveals that, compared to the direct carbon atom injection case, charge‐exchange collision is the dominant collision of neutral methane (hydrocarbons), which suppresses divertor power radiation and plasma detachment while promoting the improvement of fueling efficiency.

Radiation shielding design of the CFETR polarimeter interferometer and CO2 dispersion interferometer

A three-wave based laser polarimeter/interferometer and a CO2 laser dispersion interferometer are used to determine the electron and current density profiles on a Chinese fusion engineering test reactor (CFETR). Radiation shielding is designed for the combination of polarimeter/interferometer and CO2 dispersion interferometer. Furthermore, neutronics models of the two systems are developed based on the engineering-integrated design of CFETR polarimeter/interferometer and CO2 dispersion interferometer and the major material components of CFETR. The polarimeter/interferometer and CO2 dispersion interferometer’s neutron and photon transport simulations were performed using the Monte Carlo neutral transport code to determine the energy deposition and neutron energy spectrum of the optical mirrors. The energy depositions of the first mirrors on the polarimeter/interferometer are reduced by three orders with the whole shielding. Since the mirrors of CO2 dispersion interferometer are very close to the diagnostic first wall, shielding space is limited and the CO2 dispersion interferometer energy deposition is higher than that of the polarimeter/interferometer. The dose rate after shutdown 106 s in the back-drawer structure has been estimated to be 83 μSv h−1 when the radiation shield is filled in the diagnostic shielding modules, which is below the design threshold of 100 μSv h−1. Radiation shielding design plays a key role in successfully applying polarimeter/interferometer and CO2 dispersive interferometer in CFETR.

Outer midplane neutral density measurements and H-mode fueling studies in NSTX-U

Neutral density profiles are measured on the outboard midplane of the National Spherical Torus Experiment Upgrade (NSTX-U) using a two dimensional camera (ENDD, edge neutral density diagnostic) filtered for deuterium Balmer α (D α ) emission interpreted via simulations using the Monte Carlo neutral transport code DEGAS 2. Deuterium atomic densities n D are calculated by inverting the line-integrated D α brightness and using local measurements of electron density n e and temperature T e to determine atomic rate coefficients. The assumptions used in the derivation of n D from D α emissivity are validated using DEGAS 2 to estimate contributions to emissivity due to electron impact excitation and molecular processes. Experimental measurements and DEGAS 2 simulations are compared over a database of L- and H-mode discharges, showing good agreement in D α emissivity profiles. Residual disagreement between experiment and simulations is suggestive of the possible role of intermittent transport and uncertainties in the molecular data. DEGAS 2 simulations are further used to complement the ENDD diagnostic, extracting neutral (atomic and molecular) densities at locations where the ENDD measurement are not made or where the assumptions used in the ENDD analysis are not valid. Deuterium atomic densities and ionization profiles on the outer midplane are compared for L-mode and H-mode discharges in NSTX-U. One-way coupling of DEGAS 2 to UEDGE multi-fluid simulations constrained by experimental data is used to study edge fueling and neutral penetration in NSTX-U discharges.

Modeling of the Hα Emission from ADITYA Tokamak Plasmas

The spatial profile of Hα spectrum is regularly measured using a high-resolution multi-track spectrometer in ADITYA tokamak to study the neutral particle behavior. The Monte Carlo neutral particle transport code DEGAS2 is used to model the experimental Hα spectral emissions. Through the modeling of the spectral line profile of Hα, it is found that the neutral hydrogen, which is produced from molecular hydrogen and molecular hydrogen ion dissociation processes contributes 56% to the total Hα emission, and the atoms which are produced from charge-exchange process have 30% contribution. Furthermore, the experimentally measured spatial profile of chord integrated brightness was modeled for the two plasma discharges having relatively high and low density to understand the neutral particle penetration. The presence of neutrals inside the core region of the ADITYA tokamak is mainly due to the charge-exchange process. Furthermore, it is observed that neutral particle penetration is lower in higher density discharge.

A reduced model of neutral-plasma interactions in the edge and scrape-off-layer: Verification comparisons with kinetic Monte Carlo simulations

The 2D scrape-off-layer turbulence code (SOLT) is extended to include neutral-plasma interactions. A Boltzmann equation is derived for the evolution of the bi-normally averaged neutral distribution function, G(x,vx,t), in the radial dimension, and this evolution is included in the new scrape-off-layer turbulence code (nSOLT). Neutral-plasma interactions are mediated by charge-exchange (CX) and ionization rates based on poloidally averaged plasma density and temperature. Good agreement is obtained between asymptotically stationary neutral density profiles from nSOLT simulations and those previously obtained from the Monte Carlo neutral transport code DEGAS 2, for time-averaged NSTX H-mode plasma profiles. The sensitivity of the nSOLT neutral profiles to atomic physics parameters, with and without CX physics, is included in the comparison. In addition, nSOLT simulations that evolve the plasma in 1D, using radial diffusion as a proxy for turbulent (blob) transport, illustrate the convergence to a self-consistent neutral-plasma equilibrium sustained by a neutral source at the far-scrape-off-layer boundary and plasma heating in the core; equilibria consistent with typical NSTX Ohmic L-mode plasmas are described.

Midplane neutral density profiles in the National Spherical Torus Experiment

Atomic and molecular density data in the outer midplane of NSTX [Ono et al., Nucl. Fusion 40, 557 (2000)] are inferred from tangential camera data via a forward modeling procedure using the DEGAS 2 Monte Carlo neutral transport code. The observed Balmer-β light emission data from 17 shots during the 2010 NSTX campaign display no obvious trends with discharge parameters such as the divertor Balmer-α emission level or edge deuterium ion density. Simulations of 12 time slices in 7 of these discharges produce molecular densities near the vacuum vessel wall of 2–8 × 1017 m−3 and atomic densities ranging from 1 to 7 × 1016 m−3; neither has a clear correlation with other parameters. Validation of the technique, begun in an earlier publication, is continued with an assessment of the sensitivity of the simulated camera image and neutral densities to uncertainties in the data input to the model. The simulated camera image is sensitive to the plasma profiles and virtually nothing else. The neutral densities at the vessel wall depend most strongly on the spatial distribution of the source; simulations with a localized neutral source yield densities within a factor of two of the baseline, uniform source, case. The uncertainties in the neutral densities associated with other model inputs and assumptions are ≤50%.

Reconstruction of NSTX midplane neutral density profiles from visible imaging data

The experimental determination of neutral densities in tokamak plasmas from line radiation is only accurate in the narrow region in which both the excitation rate and neutral density are significant. We describe an alternative procedure using the DEGAS 2 Monte Carlo neutral transport code to invert light emission data obtained from a tangentially viewing camera, yielding absolute radial profiles of deuterium atoms and molecules at midplane. That the neutral source in these simulations can be adequately characterized as a uniform flux at the vacuum vessel wall is demonstrated by the similarity of the shapes of the simulated and observed brightness profiles. A second test is obtained by comparing the resulting neutral pressures at the vessel walls with data from midplane micro-ion gauges. We also show that the simulated camera image is insensitive to variations in the spatial distribution of the neutral source.

Onion skin model (OSM) analysis of EAST SOL plasmas

Two-dimensional maps of the Experimental Advanced Superconducting Tokamak (EAST) scrape-off layer (SOL) plasma conditions for ohmic, L-mode and H-mode discharges are reconstructed using an onion skin model (OSM) coupled in DIVIMP together with the Monte Carlo neutral transport code, EIRENE. The boundary conditions for OSM calculation are taken from the measurements of the Langmuir probe built into the divertor targets. The OSM-calculated values of the outboard mid-plane electron density, ne, and temperature, Te, are compared with the mid-plane measurements of ne and Te from a fast reciprocating probe. Some other characteristics of these SOL plasmas are also derived from the OSM solution, reflecting that the upstream plasma conditions are governed by the SOL collisionality to a large degree. Values of at the low-field side and the high-field side mid-plane are derived separately as a function of the distance to the separatrix for ohmic, L- and H-mode discharges, showing that increases with the distance to the separatrix at both sides and that the values of at the low-field side tends to be higher than that at the high-field side. is found to be larger than by a factor of 2–3 for all the discharges considered here. In addition, before the use of the OSM method of extracting and for EAST discharges, the reliability of this method is assessed by taking SOLPS-generated target n, T profiles as boundary conditions and by comparing the OSM-extracted cross-field transport coefficients with those input in the SOLPS modelling.

Pedestal fueling simulations with a coupled kinetic plasma–kinetic neutral transport code

The simplified neutral transport routine in the guiding center ion–electron-neutral neoclassical PIC code XGC0 has been replaced by one that calls the Monte Carlo neutral transport code DEGAS2. This provides a more realistic treatment of neutral atoms in the tokamak edge plasma and allows detailed atomic physics and plasma–material interaction processes to be incorporated into the simulations. The spatial profile of the neutral particle source used in DEGAS2 is determined from the fluxes of XGC0 ions to the material surfaces, and DEGAS2 is run in a time dependent manner synchronized with XGC0. The kinetic plasma–kinetic neutral transport capability is demonstrated with example pedestal fueling simulations.

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.

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.

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%.

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.

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.