Impurity Influx Research Articles

Sputtering Coefficients of Beryllium and Tungsten by Various Atoms from Hydrogen to Tungsten

Using computer simulation, the sputtering coefficients of Be and W targets, promising materials for the first wall and divertor in the ITER tokamak, are calculated in a wide range of incident atom energies 10–100 000 eV. The following atoms were chosen as projectiles: H, D, T, He, Be, C, N, O, Ne, Ar, W. A strong influence of the surface profile on the results obtained is shown. The limiting cases of a planar potential barrier (smooth surface) and a spherical potential barrier (rough surface) are considered. Data on the average energy and angular distributions of sputtered atoms were obtained, which are necessary for calculating the influx of impurities into the tokamak plasma. The influx of wall material atoms into the ITER tokamak plasma is estimated when the wall is sputtered by flows of fast deuterium and tritium atoms leaving the plasma.

Conceptual design of visible spectroscopy diagnostics for DTT

The project of the Visible Spectroscopy diagnostics for the Z eff radial profile measurement and for the divertor visible imaging spectroscopy, designed for the new tokamak DTT (Divertor Tokamak Test), is presented. To deal with the geometrical constraints of DTT and to minimize the diagnostics volume inside the access port, an integrated and compact solution hosting the two systems has been proposed. The Z eff radial profile will be evaluated from the Bremsstrahlung radiation measurement in the visible spectral range, acquiring light along ten Lines of Sight (LoS) in the upper part of the poloidal plane. The plasma emission will be focused on optical fibers, which will carry it to the spectroscopy laboratory. A second equipment, with a single toroidal LoS crossing the plasma centre and laying on the equatorial plane, will measure the average Z eff on a longer path, minimizing the incidental continuum spectrum contaminations by lines/bands emitted from the plasma edge. The divertor imaging system is designed to measure impurity and main gas influxes, to monitor the plasma position and kinetics of impurities, and to follow the plasma detachment evolution. The project aims at obtaining the maximum coverage of the divertor region. The collected light can be shared among different spectrometers and interferential filter devices placed outside the torus hall to easily change their setup. The system is composed of two telescopes, an upper and a lower one, allowing both a perpendicular and a tangential view of the DTT divertor region. This diagnostic offers a unique and compact solution designed to cope the demanding constraints of this next-generation tokamak fusion devices, integrating essential tools for wide-ranging impurity characterization and versatile investigation of divertor physics.

RFX-mod2 diagnostic capability enhancements for the exploration of multi-magnetic-configurations

The RFX-mod2 device, the upgraded version of the previous RFX-mod with a modified magnetic boundary, is presently under realization and will start to be operated in 2025. Significant upgrades of the diagnostic capabilities have been proposed and are under development. These include a largely increased number of in-vessel magnetic and electrostatic sensors, a new fast reciprocating manipulator for the exploration of the edge plasma in a wide range of experimental conditions, the improved Thomson scattering and soft x-ray diagnostics system for a detailed determination of the behavior of the electron temperature profile, new dedicated systems for the space and time resolved analysis of x-ray spectra and neutron rate, a reflectometric diagnostic for real-time determination of plasma position, two diagnostics devoted to the imaging of light impurities and influxes behavior along with arrays of halo current sensors. These diagnostic upgrades will be accompanied by a significant effort to improve the control of the electron density and of the impurity influxes by means of proper treatment of plasma facing components with in-vessel fixed electrodes distributed over the first wall. The described advancements will allow a deeper understanding of physics phenomena in the wide variety of magnetic configurations, including the tokamak, the reversed-field pinch and the Ultra-low q, which can be produced in RFX-mod2 thanks to its flexibility and unique MHD control capabilities.

Evolution of hydrogen isotopes retention behavior of in-situ boronization films in EAST

Effective management of hydrogen isotopes retention in plasma-facing materials (PFMs) is crucial, particularly when utilizing tritium (T) as fuel, for the success of burning plasma operations. Boronization, a widely employed technique for controlling fuel recycling and mitigating impurity influx from plasma-surface interactions into the core of burning plasma, significantly influences hydrogen isotopes retention in PFMs. In this study, boronization films were generated by ion cyclotron range of frequency (ICRF) discharge assisted boronization with C2B10H12 as boron source on tungsten substrates at Experimental Advanced Superconducting Tokamak (EAST) which employing ITER-like water-cooled W monoblock PFMs and components (PFCs), followed by in-situ glow discharge (GD) cleaning for 2 h and 20 shots (180 s) edge-plasma exposure. Employing the Material and Plasma Evaluation System (MAPES), representative samples were analyzed after each process. The resultant carbon–boron films, dense and continuous, exhibited thickness up to 120 nm and were identified as amorphous in structure. It was observed that the D2-GD cleaning resulted in a significant isotopic exchange effect, effectively reducing the hydrogen (H) retention in the carbon–boron films. This hydrogen isotope replacement efficiency was found to be influenced by the thickness of the films. Notably, after boundary plasma exposure, samples with thicker films demonstrated an enhanced capacity to capture deuterium (D), adsorbing 10 times more D than bare tungsten (W). Our findings offer transformative insights for T recycling analysis and the plasma operation of devices like International Thermonuclear Experimental Reactor (ITER), highlighting the impact of boronization and subsequent treatments on hydrogen isotope retention behavior in PFMs.

Interpretative 3D MHD modelling of deuterium SPI into a JET H-mode plasma

The pre-thermal quench (pre-TQ) dynamics of a pure deuterium ( D2 ) shattered pellet injection (SPI) into a 3MA / 7MJ JET H-mode plasma is studied via 3D non-linear MHD modelling with the JOREK code. The interpretative modelling captures the overall evolution of the measured density and radiated power. The simulations also identify the importance of the drifts of ablation plasmoids towards the tokamak low field side (LFS) and the impurities in the background plasma in fragment penetration, assimilation, radiative cooling and MHD activity in D2 SPI experiments. It is found that plasmoid drifts lead to an about 70% reduction of the central line-integrated density (compared to a simulation without drifts) in the JET D2 SPI discharge considered. Impurities that pre-exist before the SPI as well as those from possible impurity influxes related to the SPI are shown to dominate the radiation in the considered discharge. With inputs from JOREK simulations, modelling with the Lagrangian particle-based pellet code PELOTON reproduces the deviation of the SPI fragments in the direction of the major radius as observed by the fast camera. This confirms the role of rocket effects and plasmoid drifts in the considered discharge and reinforces the validity of the JOREK modelling. The limited core density rise due to plasmoid drifts and the strong radiative cooling and MHD activity with impurities (depending on their species and concentration) could limit the effectiveness of LFS D2 SPI in runaway electron avoidance and are worth considering in the design of the ITER disruption mitigation system.

Magnetic nanoparticle loading and application of weak magnetic field to reconstruct the cake layer of coagulation-ultrafiltration process to achieve efficient antifouling: Performance and mechanism analysis

Abandoning the costly development of new membrane materials and instead directly remodeling the naturally occurring cake layer constitutes a dynamic, low-cost, long-lasting, and proactive strategy to "fight fouling with fouling". Several optimization strategies, including coagulation/modified magnetic seed loading and applying a weak magnetic force (0.01T) at the ultrafiltration end, improved the anti-fouling, retention, and sieving performances of conventional ultrafiltration process during the treatment of source water having complex natural organic matter (NOMs) and small molecule micropollutants. Two modified magnetic seeds we prepared were composite nano-seed particles (Fe3O4@SiO2-NH2 (FS) and Fe3O4@SiO2@PAMAM-NH2 (FSP)). Aim of the study was to regulate the formation of cake layer via comprehensive testing of the antifouling properties of optimized processes and related mechanistic studies. It was found to be essential to enhance the interception of xanthate and tryptophan proteins in the cake layer for improving the anti-fouling performance based on the correlation and redundancy analyses, while the use of modified magnetic seeds and magnetic field showed a significant positive impact on water production. Blockage modeling demonstrated the ability to form a mature cake layer during the initial filtration stage swiftly. This mitigated the risk of irreversible fouling caused by pore blockage during the early stage of coagulation-ultrafiltration. Morphologically, the reconstructed cake layer exhibited elevated surface porosity, an internal cavity channel structure, and enhanced roughness that can promote increased water flux and retention of water impurities. These optimized the maturity of the cake layer in both time and space. Density Functional Theory (DFT), Extended Derjaguin-Landau-Verwey-Overbeek (XDLVO) theory, and Modified Extended Derjaguin-Landau-Verwey-Overbeek (MDLVO) calculations indicated aggregation behavior of matter on the cake layer to be enhanced effectively due to magnetic seed loading. This is mainly due to the strengthening of polar interactions, including hydrogen bonding, π-π* conjugation, etc., which can happen between the cake layer loaded with FSP and the organic matter. Under the influence of a magnetic field, magnetic force energy (VMF) significantly impacts the system by eliminating energy barriers. This research will provide innovative strategies for effectively purifying intricate source water through ultrafiltration while controlling membrane fouling.

Impact of divertor neutral pressure on confinement degradation of advanced tokamak scenarios at ASDEX Upgrade

Over previous campaigns, an intense experimental program on advanced tokamak (AT) scenarios, has been carried out at the ASDEX Upgrade tokamak with full-tungsten wall. These discharges have been executed shortly after the boronization of the first wall to reduce the density and the impurity influx. The confinement level of such AT discharges was found to vary considerably, even when discharges with similar, if not identical, engineering parameters were carried out. This work investigates the causes of such confinement variations. Among all plasma quantities analyzed, confinement quality of AT scenarios correlates best with divertor neutral pressure, highlighting the key role of edge and scrape-off layer physics in determining global plasma confinement. In particular, it is found that the main cause of confinement degradation is the reduction of pedestal stability, which is in turn caused by the outward shift of the maximum density gradient position typically observed when the divertor neutral pressure increases. Owing to the low density of AT discharges under analysis, the movement of the maximum density gradient position can be caused entirely by changes in deuterium outgassing from the wall, which is strongly influenced by the boron layer deposited on the plasma-facing components and by the deuterium wall inventory. Finally, the predictive capability of confinement quality with the integrated model IMEP [Luda et al., Nucl. Fusion 60, 036023 (2020)] is tested on these discharges and shows promising results.

Sputtering Yields of Beryllium and Tungsten by Various Atoms from Hydrogen to Tungsten

Sputtering yields of targets made of Be and W, promising materials for the first wall and the divertor of the ITER tokamak, are calculated in a wide range of projectile energies from 10 to 100 000 eV by computer simulation. The following atoms were chosen as bombarding particles: H, D, T, He, Be, C, N, O, Ne, Ar, and W. It is demonstrated that the shape of the surface strongly impacts the obtained results. The limiting cases of a planar potential barrier (a smooth surface) and a spherical potential barrier (a rough surface) are analyzed. The data on the average energy and angular distribution of sputtered atoms needed for calculation of the impurity influx into tokamak plasma are obtained. The influx of atoms of the wall material into the ITER tokamak plasma upon wall sputtering by fluxes of fast deuterium and tritium atoms leaving plasma is estimated.

Neoclassical transport of impurities in tokamaks with non-axisymmetric perturbations

The effect of resonant magnetic perturbations (RMPs) on neoclassical transport of impurities is calculated in DIII-D and NSTX. The neoclassical fluxes are evaluated using the NEO code with nonlinear one-fluid nonaxisymmetric equilibrium calculated using M3D-C1. Neoclassical fluxes of impurities show significant changes with RMPs if the impurity resides in low-collisionality regime, but are weakly affected by RMPs in the Pfirsch–Schlüter (P–S) regime. Charge number (Z) of the impurity affects the collisionality of impurity species, which then determines the effect of 3D fields on neoclassical impurity transport. This suggests that RMPs can be possibly used for exhaust of low-Z impurities in these facilities, but have little effect for impurities with charge number greater than 10 or so. Additionally, it is shown that it is the change in the convective fluxes which is a main contributor in modifying the neoclassical impurity fluxes in the presence of RMPs.

Schmidt number's impact during the growth of mc-silicon ingot through directional solidification furnace for photovoltaic applications: A computational investigation

The use of multi-crystalline silicon for photovoltaic solar cell applications is highly advantageous due to its low manufacturing costs. In this study, numerical simulations were conducted to investigate mass transport phenomena during the directional solidification of molten silicon for different Schmidt values at the critical Prandtl number of 0.01. The simulations were performed using the diffusion model and convection-conduction equations as the framework for the incompressible Navier-Stokes equation. The finite-volume method was used to carry out computations in a two-dimensional (2D) axisymmetric model. The diffusion flux of impurities (oxygen, carbon, and nitrogen) and dopants (phosphorus and boron) during the mc-silicon growth process was examined for Schmidt numbers Sc=0.5, Sc=5, and Sc=10. It was found that the Schmidt number Sc=0.5 is suitable for producing higher-grade mc-silicon ingots by regulating the production of impurities and dopants. The simulation results obtained in this study may provide a better understanding of the mc-silicon growth process and can be used to improve the production of high-quality mc-silicon ingots for photovoltaic applications.

Algebraic and numerical studies on the roles of momentum conservation and self-adjointness in moment-based neoclassical particle fluxes

Algebraic and numerical studies on the roles of momentum conservation and self-adjointness in moment-based neoclassical particle fluxes

Global gyrokinetic simulation for toroidal impurity mode turbulence

In this study, global gyrokinetic simulations of the toroidal impurity mode (tIM) turbulence are performed. A linear analysis shows that the tIM is an instability that occurs in the bad curvature region when the density gradients of bulk ions and impurities are in opposite directions. The tIM can be unstable even when the temperature profiles are flat. In the presence of temperature gradients, the tIM and toroidal ion temperature gradient (tITG) mode could coexist. For the small temperature gradient, the tIM is found to be dominant. The tIM turbulence drives the large inward impurity and outward ion particle transports. Furthermore, the inward ion heat flux driven by the tIM turbulence causes the ion temperature profile to be more peaked than the initial one. For the large temperature gradient, while such inward ion heat flux does not occur because of the dominant tITG mode, the large inward impurity and outward ion particle fluxes are still observed due to the subdominant tIM. These results indicate that the tIM plays an important role in turbulent heat and particle transport when impurities are injected.

Refining light impurity content estimates at the lower divertor based on experimental data in WEST

WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten (W). One of the aims of WEST is to study plasma operation with W PFCs in preparation for long-pulse operation on W divertor devices like ITER (Bucalossi et al., 2014) [1]. For long-pulse operation, the W impurity content and transport to the core plasma are critical concerns that require further measurement and interpretation in order to improve plasma performance and PFC durability. In low confinement mode (L-mode) operation, the sputtering of tungsten is not expected to be driven by the plasma itself (D), but by the main light impurities (C, B, O, N) observed at trace levels in WEST. Identifying the main W sources and understanding W transport in the SOL are key metrics for plasma performance, which are simulated with Scrape-Off Layer (SOL) impurity transport codes (SOLEDGE-ERO, OEDGE-DIVIMP). Previous studies have used O as proxy for light impurity with a global content varying from 3 to 5% in SOL transport simulations (Genova et al., 2021; Rooij et al., 2020; Klepper et al., 2022) [2–4]. A power scan was performed in a standard WEST magnetic equilibrium with LH injected power up to 5 MW. Density and temperature were measured with flush-mounted Langmuir probes (Dejarnac et al., 2021) [5]. The impurity fluxes are estimated by applying the number of ionizations per photons (S/XB) coefficients calculated with the collisional-radiative model “ColRadPy” (Johnson et al., 2019) [6] to the spectral radiance measured with the WEST visible spectroscopy system (Meyer et al., 2016) [7]. This work details how the low-Z impurity content (C, O) near the divertor target was refined using the visible spectroscopy for several impurity charge states, coupled with ColRadPy to infer the content of all impurity charge states at the divertor. First results shows that the main light impurity is C and not O as originally assumed. This inferred impurity content is then used to evaluate the W sputtering at the divertor, considering the sheath at the divertor. A quantitative comparison is performed with the W flux at the divertor measured with visible spectroscopy that shows a qualitative agreement, but not fully consistent at the strike point position. This, in turn, suggests that the impurity content is not complete with O and C only.

First feedback-controlled divertor detachment in W7-X: Experience from TDU operation and prospects for operation with actively cooled divertor

In the last experimental campaign (OP1.2b) of the stellarator Wendelstein 7-X (W7-X), boronisation as a mean for first wall conditioning was applied for the first time which led to strongly reduced impurity fluxes from plasma-facing components. Thermal detachment at the uncooled target plates of the test divertor unit (TDU) was reached at higher plasma densities and was accompanied by high recycling of neutrals at the target plate (Feng et al., 2021), (Jakubowski et al., 2021) [1] . A feedback control system was established in W7-X to actively control the gas injection (actuator) for plasma fuelling and impurity seeding (Effenberg et al., 2019) [2] through the divertors. It allowed very successful stabilisation of the detached plasma condition state as well as mitigation of thermal overloads to some baffle tiles. Different routinely available diagnostic signals were used as input parameters (sensors). We describe the setup of the feedback control system, its performance and provide some example results with the main focus on the development of the control scheme which led to the detachment stabilisation over the entire longest (30 s) high-power discharge at W7-X so far. In view of the achieved very successful detachment stabilisation and the necessity to include simultaneous optimisation of the core performance in the future, the feedback control system is being upgraded for the upcoming campaign (OP2.1) in which the water cooled and hereby inherently steady-state capable divertor has been currently installed. The prospects and some experiment ideas for active detachment control are discussed.

Analytical model for the combined effects of rotation and collisionality on neoclassical impurity transport

The influence of rotation, collisionality and trapped particle fraction on the magnitude and direction of neoclassical impurity transport in tokamaks is analyzed using an extensive database of drift-kinetic simulations with the NEO code. It is shown that an operational window opens at sufficiently high Mach number and low collisionality, where the magnitude of the temperature screening of impurities increases with higher rotation. If the collisionality increases, this effect is quickly lost and the temperature gradient then drives an inward impurity flux when rotation is present. The boundary between these two regimes is calculated as a function of the trapped particle fraction, and it is shown that plasma parameters achieved in recent JET experiments allow them to access the new beneficial regime, in accordance with observations of reduced tungsten accumulation. Applications to ASDEX Upgrade experiments where these effects become relevant are also presented, and the implications for ITER are discussed. A method for extracting the physically distinct Pfirsch–Schlüter (PS) and banana-plateau (BP) neoclassical flux components from the NEO output is introduced and employed to construct a model that describes them analytically at arbitrary rotation and collisionality. The beneficial behavior of the screening with rotation is found to be a BP effect, in contrast to the known detrimental role of rotation in the PS component. The new analytical model is able to reproduce the results of NEO when modeling radial profiles of transport coefficients from experimental kinetic profiles, with the added feature of isolating the BP and PS components for additional physical analysis, while remaining well suited for fast applications.

Development of ECRH-based methods for assisted discharge burn-through: Experiment and simulation

Electron Cyclotron (EC) waves will be routinely used in future reactors not only for plasma heating and/or non-inductive current drive during the flat top but also to assist the plasma start-up phase in large tokamaks with superconductive coils. In ITER, for example, EC start-up is foreseen since first plasma operation. To limit the level of stray radiation, ECRH can be used after ohmic breakdown, as a robust solution to successfully sustain the plasma burn-through in the presence of pre-filling gas and impurity influx from the wall. On ASDEX Upgrade (AUG), a series of dedicated experiments have been performed using EC heating (X2) with a controlled Ne impurity injection in the prefill phase, to mimic non-favourable burn-through conditions such as would be expected in a discharge following a disruption event. The time for EC heating onset has been optimised to assist the early burn-through and a scan of the Ne concentration has been performed to find the threshold for successful burn-through conditions for two ECH power levels (0.7 and 1.4 MW). The toroidal magnetic field flexibility has been also documented, with the cold resonance position being shifted up to 13% in major radius to match the ITER condition. These experiments showed that optimised settings of ECH power (onset and duration of the pulse) have a key role in making feasible the early Ne burn-through (with Ne concentration up to 14% and EC power of 1.4 MW). Successful pulses will be extended to study stationarity and clean up properties. For an efficient and robust use of such a technique, it is essential to develop appropriate models capable of describing present experiments and of extrapolating (or predicting) to future scenarios. In this work, the predictive 0D model for the burn-though phase BKD0 [1] has been used to reproduce experimental results and estimate the power required for a successful burn-through as a function of the impurity concentration, finding that ECH power of 1.4 MW is required to sustain burn-through with more than 20% of Ne. The scalability of the model has been also tested on TCV [2] and its implication for ITER will be discussed.

Models of recrystallization activated by a diffusion flow of impurities from a thin-film coating with a convection term at the crystal surface: exact solutions

Two models of recrystallization are proposed taking into account the convective flux of impurity exchange between the polycrystalline and the thin-film coating. The special boundary modes of recrystallization described by the single-phase and two-phase Stefan problems with the boundary condition at coated surface containing the convective term. The exact solutions of the formulated problems corresponding to the grain-boundary concentration of impurities are obtained. The detail theoretical analysis focused on the third type problem shows that the concentration of impurities and the width of the recrystallized layer increase with an increase in the annealing time. An increase in intensity of impurity exchange between the polycrystalline and the coating promotes an increase in the width of the recrystallized layer. The recrystallization front position increases with an increase in the surface concentration of impurities and it decreases with an increase in the intensity of the impurity flux from the surface. The rate of recrystallization kinetics increases with an increase in the intensity of impurity exchange between the polycrystalline and the coating.

Peripheral temperature gradient screening of high-Z impurities in optimised ‘hybrid’ scenario H-mode plasmas in JET-ILW

Screening of high-Z (W) impurities from the confined plasma by the temperature gradient at the plasma periphery of fusion-grade H-mode plasmas has been demonstrated in the JET-ILW (ITER-like wall) tokamak. Through careful optimisation of the hybrid-scenario, deuterium plasmas with sufficient heating power ( 32 MW), high enough ion temperature gradients at the H-mode pedestal top can be achieved for the collisional, neo-classical convection of the W impurities to be directed outwards, expelling them from the confined plasma. Measurements of the W impurity fluxes between and during edge-localised modes (ELMs) based on fast bolometry measurements show that in such plasmas there is a net efflux (loss) between ELMs but that ELMs often allow some W back into the confined plasma. Provided steady, high-power heating is maintained, this mechanism allows such plasmas to sustain high performance, with an average D–D neutron rate of s−1 over a period of ∼3 s, after an initial overshoot (equivalent to a D–T fusion power of ∼9.4 MW), without an uncontrolled rise in W impurity radiation, giving added confidence that impurity screening by the pedestal may also occur in ITER, as has previously been predicted (Dux et al 2017 Nucl. Mater. Energy 12 28–35).

Low-Temperature Epitaxial Growth by Quiescent Plasma-Enhanced Chemical Vapor Deposition at Atmospheric Pressure

Atmospheric pressure (AP) plasma provides an alternative approach to low-cost thin-film deposition. The low throughput of epitaxial growth, which limits productivity in semiconductor manufacturing, can be addressed by using AP plasma. In principle, AP plasma does no damage and enables local heating of the deposition surface. In Si epitaxial growth using AP plasma, hydrogen incorporation and quiescent gas flow are shown to be key factors controlling epitaxial growth even under a high environmental impurity flux and at a low substrate temperature of 150 °C. Quiescent plasma at AP is promising for future epitaxial processing owing to its low cost and high productivity.

Development of an Integrated Multidiagnostic to Assess the High-Z Impurity Fluxes in the Metallic Environment of WEST Using IMAS

WEST is an actively cooled, long-pulse tokamak with nearly all plasma-facing components (PFC) made of tungsten. One of the aims of WEST is to study plasma operations with tungsten PFCs in preparation for long-pulse operations on high-Z divertor devices, such as ITER. For long-pulse operation, the high-Z impurity content and transport to the core plasma are critical concerns that require further measurement and interpretation to improve plasma performance and PFC durability. This work details the impurity influxes in WEST during a series of discharges in which the lower hybrid (LH) injected power was incrementally increased. An analysis has been performed of measurements collected from an array of edge diagnostics. Visible spectroscopy was utilized to measure the spectral radiances generated by fuel particles ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$D$ </tex-math></inline-formula> ) and impurities ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$W$ </tex-math></inline-formula> , <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$O$ </tex-math></inline-formula> , and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$C$ </tex-math></inline-formula> ) at the divertor and at the antennas with a newly developed spectral peak-fitting tool used to analyze the data in WEST. The scrape-off layer (SOL) plasma conditions ( <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$n_{e}$ </tex-math></inline-formula> and <inline-formula xmlns:mml="http://www.w3.org/1998/Math/MathML" xmlns:xlink="http://www.w3.org/1999/xlink"> <tex-math notation="LaTeX">$T_{e}$ </tex-math></inline-formula> ) measured at the divertor target with flush Langmuir probes and near the outer mid-plane (OMP) using reciprocating Langmuir probes (RCPs) are used to evaluate the number of ionizations per photons (S/XB) coefficients required to estimate the impurity fluxes obtained with the collisional-radiative model ColRadPy. The array of edge diagnostics discussed in this work, coupled with SOL plasma modeling tools, represents a multidiagnostic interpretative modeling workflow that will continue to be applied to upcoming experimental campaigns on the WEST experiment to assess high-Z impurity transport.