Divertor Detachment Research Articles

Spatial Profiles of NeVI-NeX Emission in ECR-Heated Discharges of the Large Helical Device with Divertor Detachment Induced by RMP Application and Ne Impurity Seeding

Spatial Profiles of NeVI-NeX Emission in ECR-Heated Discharges of the Large Helical Device with Divertor Detachment Induced by RMP Application and Ne Impurity Seeding

Progress from ASDEX Upgrade experiments in preparing the physics basis of ITER operation and DEMO scenario development

An overview of recent results obtained at the tokamak ASDEX Upgrade (AUG) is given. A work flow for predictive profile modelling of AUG discharges was established which is able to reproduce experimental H-mode plasma profiles based on engineering parameters only. In the plasma center, theoretical predictions on plasma current redistribution by a dynamo effect were confirmed experimentally. For core transport, the stabilizing effect of fast ion distributions on turbulent transport is shown to be important to explain the core isotope effect and improves the description of hollow low-Z impurity profiles. The L–H power threshold of hydrogen plasmas is not affected by small helium admixtures and it increases continuously from the deuterium to the hydrogen level when the hydrogen concentration is raised from 0 to 100%. One focus of recent campaigns was the search for a fusion relevant integrated plasma scenario without large edge localised modes (ELMs). Results from six different ELM-free confinement regimes are compared with respect to reactor relevance: ELM suppression by magnetic perturbation coils could be attributed to toroidally asymmetric turbulent fluctuations in the vicinity of the separatrix. Stable improved confinement mode plasma phases with a detached inner divertor were obtained using a feedback control of the plasma β. The enhanced D α H-mode regime was extended to higher heating power by feedback controlled radiative cooling with argon. The quasi-coherent exhaust regime was developed into an integrated scenario at high heating power and energy confinement, with a detached divertor and without large ELMs. Small ELMs close to the separatrix lead to peeling-ballooning stability and quasi continuous power exhaust. Helium beam density fluctuation measurements confirm that transport close to the separatrix is important to achieve the different ELM-free regimes. Based on separatrix plasma parameters and interchange-drift-Alfvén turbulence, an analytic model was derived that reproduces the experimentally found important operational boundaries of the density limit and between L- and H-mode confinement. Feedback control for the X-point radiator (XPR) position was established as an important element for divertor detachment control. Stable and detached ELM-free phases with H-mode confinement quality were obtained when the XPR was moved 10 cm above the X-point. Investigations of the plasma in the future flexible snow-flake divertor of AUG by means of first SOLPS-ITER simulations with drifts activated predict beneficial detachment properties and the activation of an additional strike point by the drifts.

Confinement improvement during detached phase with RMP application in deuterium plasmas of LHD

In order to explore the compatibility of good core plasma performance with divertor heat load mitigation, the interaction between cold edge plasma and core plasma transport, including the edge transport barrier (ETB), has been analysed in the divertor detachment discharges of deuterium plasmas in LHD with resonant magnetic perturbation (RMP) field application. The RMP application introduces a widened edge stochastic layer and sharp boundary in the magnetic field structure between the confinement region and the edge stochastic layer. The widened edge stochastic layer enhances impurity radiation and provides stable detachment operation as compared with the case without RMP. It is found that ETB is formed at the confinement boundary at the onset of detachment transition. However, as the detachment deepens, the resistive pressure gradient-driven MHD mode is excited, which degrades the ETB. At the same time, however, the core transport decreases to keep global plasma stored energy (W p) unchanged, showing clear core-edge coupling. After a gradual increase of density fluctuation during the MHD activity, a spontaneous increase of W p and the recovery of ETB are observed while the detachment is maintained. Then, the coherent MHD mode ceases and ELM-like bursts appear. In the improved mode, impurity decontamination occurs, and the divertor heat load increases slightly. Key controlling physical processes in the interplay between core and cold edge plasma are discussed. A comparison between deuterium and hydrogen plasmas shows that hydrogen plasmas exhibit similar features to the deuterium ones in terms of density and magnetic fluctuations, impurity decontamination towards higher confinement, etc. But most of the features are modest in the hydrogen plasmas and thus no clear confinement mode transition with clear ETB formation is defined. Better global confinement is obtained in the deuterium plasmas than the hydrogen ones at a higher radiation level.

Recent results from deuterium experiments on the large helical device and their contribution to fusion reactor development

In recent deuterium experiments on the large helical device (LHD), we succeeded in expanding the temperature domain to higher regions for both electron and ion temperatures. Suppression of the energetic particle driven resistive interchange mode (EIC) by a moderate electron temperature increase is a key technique to extend the high temperature domain of LHD plasmas. We found a clear isotope effect in the formation of the internal transport barrier in high temperature plasmas. A new technique to measure the hydrogen isotope fraction was developed in the LHD in order to investigate the behavior of the isotope mixing state. The technique revealed that the non-mixing and the mixing states of hydrogen isotopes can be realized in plasmas. In deuterium plasmas, we also succeeded in simultaneously realizing the formation of the edge transport barrier (ETB) and the divertor detachment. It is found that resonant magnetic perturbation plays an important role in the simultaneous formation of the ETB and the detachment. Contributions to fusion reactor development from the engineering point of view, i.e. negative-ion based neutral beam injector research and the mass balance study of tritium, are also discussed.

Conceptual design of reciprocating probes and material-testing manipulator for tokamak COMPASS Upgrade

Three new in-vessel manipulators are designed and built for the new COMPASS Upgrade tokamak with uniquely high vessel temperature (250–500 °C) and heat flux density (perpendicular to divertor surface q ⊥ ∼ 80 MW/m2 and q ∥ ∼ GW/m2 at separatrix), which challenges the edge plasma diagnostics. Here we show their detailed engineering designs supported by heat conduction and mechanical models. Deep reciprocation of electrostatic probes near the separatrix should be possible by optimizing older concepts in (a) the head and probe geometry, (b) strongly increasing the deceleration up to 100× gravity by springs and strengthening the manipulator mechanical structure. One reciprocates close to the region of edge plasma influx (the outer midplane), the other at the plasma sink (between the outer divertor strike point and X-point), for studying the plasma divertor (impurity-seeded) detachment and liquid metal vapor transport. Both probe heads are equipped with a set of ball-pen and Langmuir probes, measuring reliably and extremely fast (10−6 s) local (1 mm resolution) plasma potential, density, electron temperature and heat flux and even ion temperature with 10−5 s resolution. The divertor manipulator (without reciprocation) will place various material test targets at the outer divertor. Unique will be its capability to increase 15× the surface heat flux with respect to the surrounding tungsten tiles just by controllable surface inclination of the test targets. We plan to test liquid metal targets where such inclined surface was found critical to achieve the desired mode with lithium vapor shielding. Even in the conservative expected performance of COMPASS Upgrade, we predict to reach and survive the EU DEMO relevant heat fluxes.

Numerical studies of the influence of seeding locations on D-SOL plasmas in EAST

Impurity seeding has been found effective for divertor detachment operations and the seeding location plays a key role in this process. In this work, we use the fluid code SOLPS-ITER to study the influence of seeding locations on divertor and scrape-off layer (D-SOL) plasmas in Experimental Advanced Superconducting Tokamak (EAST) with neon seeding. Simulation results indicate that the neon is a highly effective impurity in mitigating the heat flux and electron temperature peaks on the target of the divertor and achieving the partial detachment on both inner and outer targets. Further, by comparing results of the seeding at the private-flux region (PFR) plate (called ‘TP’ location) and the outer target (called ‘XP’ location), we find that the impurity density and power radiation for TP case are higher in core and upstream regions and lower in the divertor region than that for seeding at the XP, and the difference becomes more and more obvious as the seeding rate increases. It clearly demonstrates that the seeding at the XP location is more appropriate than at the TP location, especially in high seeding rate conditions.

Divertor detachment in the pre-fusion power operation phase in ITER during application of resonant magnetic perturbations

Detachment of the divertor plasma during application of resonant magnetic perturbation (RMP) fields is evaluated for hydrogen H-mode plasma during the first pre-fusion power operation (PFPO-1) phase in ITER by 3D plasma boundary modelling with EMC3–EIRENE. Plasma response effects from a linearized, resistive, single fluid MHD model are discussed, which includes partial screening of the externally applied field—but also field amplification near the separatrix. This field amplification is found to play a pivotal role for the magnetic footprint on the divertor targets, but is sensitive to model parameters. Extensions of the footprint beyond the straight portions of the ITER vertical divertor targets, optimized for high stationary heat flux handling, may be possible depending on the level of toroidal rotation in the plasma. Exhaust from the bulk plasma is guided by the helical corrugations (lobes) of the perturbed separatrix, and this results in an upstream heat flux that is distributed over these lobes with lower peak values than in the typical radial heat flux profiles seen in the absence of magnetic perturbations. As a consequence, an earlier onset (with respect to the upstream density) of detachment is found in the traditional strike zone when RMPs are applied, but secondary, non-axisymmetric strike locations appear—and those remain attached at temperatures above 10 eV. Neon seeding can mitigate these non-axisymmetric heat loads, but this becomes less efficient for large magnetic footprints.

Steady-state sustainment of divertor detachment with multi-species impurity seeding in LHD

Multi-species impurity seeding is an advanced operation scenario to mitigate the divertor heat load for the realization of future fusion reactors. In the large helical device, divertor detachment is successfully sustained using higher-Z (krypton, Kr) and lower-Z (neon, Ne) superimposed seeding. Emission from Kr impurities is drastically enhanced if it is followed by Ne seeding. Plasma radiation can be enhanced even at the upstream region in the edge plasma compared with Ne only seeded plasmas with suppression of impurity accumulation toward the central plasma. The characteristics of divertor heat load reduction and energy confinement are comparable between the Kr + Ne seeding and Ne only seeding under the same radiation fraction. However, while the detachment in Ne only seeding is transient, the detachment in Kr + Ne seeding is stable. It indicates that multi-species impurity seeding can be competitive for steady-state operation although further investigation is desired about the balance between divertor heat load reduction, impurity screening, and confinement degradation. The Kr emission enhancement is strongly affected by electron density and temperature at the last closed flux surface resulting in impurity penetration.

Detachment in conventional and advanced double-null plasmas in TCV

Divertor detachment is investigated on the Tokamak à Configuration Variable (TCV) for conventional and alternative double-null (DN) magnetic geometries in L-mode, Ohmic density ramps and then compared to precisely matched single-null (SN) geometries. The poloidal flux expansion at the outer strikepoint(s) is varied by a factor 2 and the total flux expansion by 30% in lower single null (LSN) and DN configurations. Leg power sharing for DN and near-DN in attached conditions is in line with previous studies in other devices, with balanced power partition between upper and lower divertor achieved for |δR sep| ≲ λ q /2. This results in a reduction in the lower outer target integral and peak heat flux compared to a LSN by 30%. Unlike previous studies, λ q is fairly insensitive to δR sep. The detachment threshold in these geometries is investigated from target measurements with wall-embedded Langmuir probes and two-dimensional CIII line emissivity profiles across the two divertor regions. DN plasmas display clear benefits compared to their LSN counterparts. Transitioning from a LSN to DN shows a substantial reduction, of approx. 20%–25%, in the detachment threshold for the active divertor legs as well as a 50% higher radiated fraction at all . The density limit is simultaneously reduced in DN by 10%–20%. Across the explored range, poloidal flux expansion has only a small effect on the detachment threshold in LSN (as seen in previous experiments), and no effect in DN, with similar observations for the total flux expansion. In general, no strong benefits of increased poloidal or total flux expansion are observed across the explored range.

PHAD: a phase-oriented disruption prediction strategy for avoidance, prevention, and mitigation in JET

The ideal operational scenario for the future tokamak reactor is disruption-free operation. However, so far all the experimental evidence indicates that disruptions are unavoidable and can occur with alarming frequency when approaching reactor conditions (low q 95, high radiated fraction, divertor detachment, etc). In this article, a unified strategy for disruption avoidance, prevention, and mitigation is proposed and validated on JET data. The approach is based on three phase-oriented predictors to detect the main instabilities leading to the undesired and sudden end of the discharge. The first model detects dangerous profiles as an early indication of a critical situation. The second is designed to identify multifaceted asymmetric radiation from the edge and other abnormal radiative events. The third model is devoted to mitigation, and triggers alarms around few tens of ms before the beginning of the current quench. The models have been trained and tested with a database of almost 1000 JET discharges of recent campaigns with the ITER-like wall. The overall performances are very close to 100% of successful detections with a few percent of false alarms. In addition to the first systematic use of visible cameras for disruption prevention in JET, the most relevant aspect of this work is related to the distribution of the alarms of the three predictors, which do not overlap and are sequential. Consequently, the three predictors are meant to work in parallel over running discharges and, depending on which one triggers the alarm, the cause can be determined and approximate remaining time to intervene can be estimated, potentially allowing for the optimisation of the remedial actions.

Numerical assessment of the new V-shape small-angle slot divertor on DIII-D

The small-angle slot (SAS) divertor of the DIII-D tokamak, and its upcoming upgrade, the V-shape small-angle slot (SAS-V) divertor, are numerically investigated using the SOLPS-ITER code package, including the effect of particle drifts, for a range of plasma density, heating power, strike point position in the slot, and for both magnetic field directions. The simulations show that the electron temperature near the strike point is reduced in SAS-V compared to SAS, for both magnetic field directions, such that SAS-V achieves divertor detachment at a lower value of the outboard mid-plane separatrix electron density. The detachment threshold is lower because the V-shape focuses recycling neutrals on the V-end, densifying and cooling the plasma in the slot. At sufficiently high density, the V-shape also reduces the radial gradient of the temperature profile at the target, which in turns reduces the radial electric field and the E × B drift velocities, further densifying and cooling the plasma in the slot and leading to detachment. The V-shape effect, however, is reduced for higher heating power. With more heating power, the detachment density increases, reducing the ionization mean free path of recycled neutrals, which therefore become less sensitive to target shape changes. This suggests that in a fusion reactor, where the heating power is high, optimization of the divertor target shape needs to be combined with other strategies to lower the detachment density, such as in-slot injection of low-Z impurities.

The role of plasma-molecule interactions on power and particle balance during detachment on the TCV tokamak

This paper shows experimental results from the TCV tokamak that indicate plasma-molecule interactions involving and possibly D− play an important role as sinks of energy (through hydrogenic radiation as well as dissociation) and particles (ions) during divertor detachment if low target temperatures (<3 eV) are achieved. Both molecular activated recombination (MAR) and ion source reduction due to a power limitation effect are shown to be important in reducing the ion target flux during a density ramp. In contrast, the electron–ion recombination (EIR) ion sink is too small to play an important role in reducing the ion target flux. MAR or EIR do not occur during N2 seeding induced detachment as the target temperatures are not sufficiently low. The impact of is shown to be underestimated in present (vibrationally unresolved) SOLPS-ITER simulations, which could result from an underestimated rate. The converged SOLPS-ITER simulations are post-processed with alternative reaction rates, resulting in considerable contributions of to particle and power losses as well as dissociation below the D2 dissociation area. Those findings are in quantitative agreement with the experimental results.

High-performance ECRH at W7-X: experience and perspectives

The second operation phase of W7-X (OP1.2) showed the potential of exclusively electron cyclotron resonance heating (ECRH)-sustained plasma operations in stellarators. Employing multi-pass ECRH scenario in the second harmonic O-mode (O2-ECRH), stationary densities of up to 1.4 × 1020 m−3 could be achieved. This scenario also made stationary divertor detachment possible, which is a reactor-relevant scenario for power and particle exhaust. At high densities and with sufficiently high density gradients for an improved ion confinement, the coupling between the electrons and ions was strong enough to bring the ion temperature to values above 3 keV and to the neoclassical limit for some magnetic configurations, thus enabling to test the W7-X neoclassical optimization. The planned enhancement of the ECRH performance will enable to advance towards reactor-relevant beta values and to investigate their stability and confinement of fast particles, which is a priority goal of W7-X.

Dependence of upstream SOL density shoulder on divertor neutral pressure observed in L-mode and H-mode plasmas in the EAST superconducting tokamak

Upstream density profiles in the scrape-off layer (SOL) have been examined in low-confinement mode (L-mode) and high-confinement mode (H-mode) plasmas in the EAST superconducting tokamak. A weak density shoulder forms in the near SOL region in upper single-null configurations when the neutral pressure measured at the lower divertor exceeds a threshold value of 2 × 10−2 Pa in L-mode plasmas. When the neutral pressure is below this threshold, the weak density shoulder is absent and the sidebands of the lower hybrid waves associated with SOL parametric instabilities are reduced. Active detachment control with neon–deuterium seeding demonstrate that the weak density shoulder can form before the onset of the outer divertor detachment as long as the neutral pressure is above the threshold. Furthermore, no remarkable expansion of a shoulder is observed during divertor detachment, suggesting that divertor detachment is not a necessary condition for the formation or growth of a density shoulder. Through the increase in neutral pressure in the lower divertor by an order of magnitude, the weak shoulder was observed to expand into the far SOL and reach the leading edge of the limiter. The results in L-mode discharges identified the neutral pressure in the lower divertor as a primary factor for the formation of an SOL density shoulder in the upper single-null discharges. For the type-I ELMy H-mode plasmas, a similar density shoulder was detected during the inter-ELM phase when the neutral pressure in the lower divertor exceeded a threshold value of 4 × 10−2 Pa. On the other hand, the shoulder was absent when the divertor neutral pressure went below this threshold even though the plasma discharge was conducted with a higher core line-averaged density and divertor collisionality. This is consistent with the observations in L-mode plasmas. The neutral particle ionization of the working gas is thus believed to play a key role during the formation of the SOL density shoulder in the EAST tokamak.

Experimental study of the influence of gas puff locations on H-mode boundary plasmas with argon seeding on EAST

To investigate the optimal scenario of impurity seeding to obtain divertor plasma detachment for target protection, experiments with Ar&D2 seeding from two different poloidal locations, the upper outer (UO) divertor target and lower outer (LO) target, were carried out on EAST. Partial energy detachment (the electron temperature near the strike point T e,spt ⩽ 10 eV) were obtained with Ar&D2 mixture puffing from the UO target and, for the first time, from the LO target into H-mode plasmas in the upper single null (USN) configuration. The peak heat flux q t on the UO target was significantly reduced (by ∼80%). The rollover of ion flux density j s did not appear probably due to insufficient momentum loss, which is independent of the puff locations. The poloidal asymmetries of particle and heat fluxes on the targets have also been investigated. The UO-dominant asymmetry of particle flux was reversed, while the UO-dominant asymmetry of heat flux was mitigated but not reversed. The plasma confinement dropped by only 14% and 8.2% in the LO-puff case and UO-puff case, respectively, during detachment. The high level of C may contributed greatly to the higher radiation in the bulk plasma region and the greater decline in W MHD in the LO-puff case.

Development of an 11-channel multi wavelength imaging diagnostic for divertor plasmas in MAST Upgrade.

Divertor detachment and alternative divertor magnetic geometries are predicted to be promising approaches to handle the power exhaust of future fusion devices. In order to understand the detachment process caused by volumetric losses in alternative divertor magnetic geometries, a Multi-Wavelength Imaging (MWI) diagnostic has recently been designed and built for the Mega Amp Spherical Tokamak Upgrade. The MWI diagnostic will simultaneously capture 11 spectrally filtered images of the visible light emitted from divertor plasmas and provide crucial knowledge for the interpretation of observations and modeling efforts. This paper presents the optical design, mechanical design, hardware, and test results of an 11-channel MWI system with a field of view of 40°. The optical design shows better than 5mm FWHM spatial resolution at the plasma on all 11 channels across the whole field of view. The spread of angle of incidence on the surface of each filter is also analyzed to inform the bandwidth specification of the interference filters. The results of the initial laboratory tests demonstrate that a spatial resolution of better than 5mm FWHM is achieved for all 11 channels, meeting the specifications required for accurate tomography.

Observation of fully detached divertor integrated with improved core confinement for tokamak fusion plasmas

Integration of divertor detachment with a high-performance (βN ∼ 3, βp &amp;gt; 2, H98 ∼ 1.5) core plasma has been demonstrated in DIII-D high-βp (poloidal beta) plasmas associated with a sustained core internal transport barrier (ITB) and an H-mode edge transport barrier (ETB). Such good core-edge integration has been achieved for both neon and nitrogen seeding, for both favorable and unfavorable B-field directions, independently from the impurity puffing locations, though these variations play important roles on divertor characteristics. Compared to the standard H-mode plasmas, the high-βp plasma exhibits a much wider window of detachment compatible with high confinement core. Fully detached divertor plasmas with low plasma temperature (Te &amp;lt; 5 eV), low particle flux, and low heat flux across the entire divertor target plate were obtained by using nitrogen seeding. This detached high-βp plasma is compatible with a newly developed detachment control system which can help optimize the nitrogen gas flow rate. Several features, i.e., the high edge safety factor in the high-βp scenario, impurity injection, closed divertor and reduced heating power requirement due to the high confinement, facilitate the achievement of full divertor detachment at lower density. Instead of degrading global performance, the divertor detachment facilitates the access to an even stronger ITB at large radius with a relatively weak ETB through self-organized synergy between ITB and ETB, leading to sustained high confinement. The strengthening of the large-radius ITB compensates for the ETB degradation associated with divertor detachment. In addition, a weak ETB naturally has smaller edge localized modes (ELMs). In particular, with neon injection, a long-period no-ELM H-mode phase has been achieved simultaneously with high-performance core and partially detached divertor plasmas. These results demonstrate the possibility of integrating excellent core plasma performance with an effective divertor solution, an essential step toward steady-state operation of reactor-grade plasmas.

Effects of divertor electrical drifts on particle distribution and detachment near the divertor target plate in DIII-D

Strong impacts of drifts on the divertor plasma in–out asymmetry and detachment are demonstrated in DIII-D with an open divertor configuration. For forward toroidal field, BT, i.e., with the ion B × ∇B drift toward the divertor, the particle flux to the inner divertor, as represented by the Langmuir probe measured ion saturation current (Jsat), exhibits a double peak structure, with electron temperature, lower at the inner target. Reversing the BT direction reverses both the radial and poloidal E × B flows, leading to a broad particle flux profile in the outboard scrape-off layer (SOL) with a similar double-peak structure to that observed at the inner target with forward BT. The correlation of a double peak structure with divertor temperature profiles confirms physical coupling between the drift flow and sheath boundary condition and their strong impact on divertor profiles. In addition, under reversed BT conditions, increasing the density flattens the target temperature profile. However, Jsat remains high away from the strike point, rendering it difficult to achieve an “effective” detached plasma, i.e., with effective reduction in both peak heat flux and peak temperature (in the far SOL). In contrast, divertor detachment with a cold and flat temperature profile can be achieved at both target plates with the forward BT.

Nuclear analysis of the DEMO divertor survey visible high-resolution spectrometer

Spectroscopic measurements have been recently proposed in DEMO for divertor detachment control. In its current design, the DEMO Divertor Survey Visible High-Resolution Spectrometer is foreseen to perform spectroscopy measurements by integrating three optical subsystems into an equatorial port (EP). Behind the first wall, light travels through a set of metallic mirrors and ducts before it reaches the closure plate of the EP. This paper presents a nuclear analysis performed with the Monte Carlo simulation program MCNP6 for two alternative configurations of the system. The results show that the configuration with 5 mirrors per transmission line is very effective to reduce the neutron streaming through the port. However, it will not be possible, with the current design, to introduce standard electronics along the spectroscopy ducts, as the dose rate limits for non-critical electronic components are exceeded in both configurations. In the plasma-facing mirrors, the heat loads are below 2 mW/cm3, which shows that the strategy of recessing the first mirrors and placing them behind small-diameter openings is effective to decrease the loads in the mirrors. FISPACT simulations for different materials show that material transmutation in the mirrors will be negligible throughout the DEMO reactor lifetime.

An analysis of controlled detachment by seeding various impurity species in high performance scenarios on DIII-D and EAST

Experiments at DIII-D and EAST have demonstrated simultaneous high confinement, divertor detachment, and active control of detachment level, all of which are required for ITER. Comparing detachment control via Te and Jsat, it appears that Te is the most straightforward sensor to use for accessing detachment onset, while Jsat offers more precise control of degree of detachment. Based on these results, control using nitrogen seeding has so far shown the best ability to follow a target value with the low disruptivity and little to no degradation of performance when an Internal Transport Barrier (ITB) is present, but not all facilities allow its use. Neon seeding also can be paired with feedback control with low impact on core performance as long as there is no disruption, however shots with neon seeding commonly disrupted during these experiments. Argon is effective in EAST, but tends to degrade performance (by ≈10%βp) when detachment is achieved. With ideal conditions and strike point position control, data from a single Langmuir probe are an acceptable input to the control algorithm, but this simple system is easily defeated by strike point displacement comparable to the Te or Jsat scale lengths. The presence of an ITB seems to be critical to retaining core performance in detachment in these parameter ranges, as the pedestal pressure tends to decrease as a result of impurity seeding.