dataesr/scientific-paragraphs-categorization huggingface dataset:
g. the well-known Lengyel model  [76]  which generally assume c imp = constant along a given flux tube). A detailed comparison with Lengyel for the case of Ne-seeding  [77]  in fact finds that the simple model is conservative in comparison with the SOLPS database and predicts c imp in the range 4-10 times higher at the point of ion flux rollover at the OT (see also Section 3. 2. 2). On the basis of this comparison, it would seem that both N and Ne would be acceptable divertor radiators for ITER, unlike in most current devices with metal PFCs. This question has also been discussed in a companion paper  [78]  through the medium of a pure simulation study using the SOLPS-ITER code in which the behaviour of N and Ne seeded H-modes on ASDEX Upgrade and ITER (both characterized by vertical target configurations) is examined.
The study differs from that of the SOLPS-4. 3 database in having drifts activated and including an H-mode pedestal region (see also Section 3. 2. 2). In the case of ITER, the inclusion of drifts does not radically change the distribution of radiation nor the target loading found in SOLPS-4. 3 for either N or Ne at the higher values of p n (where operation at high performance is most likely), as might be expected given the high toroidal field in ITER and hence the lower drift strength. On ASDEX Upgrade, drifts are in contrast seen to have a very strong impact on the HFS divertor region, cooling the plasma and increasing the impurity concentration there. As proposed in detail in  [79]  (from which the ASDEX Upgrade part of the study in  [78]  was extracted), leakage of impurities to the main chamber SOL can then occur with more or less efficiency via frictional coupling to the main ion flows depending on the location of the ionization position of the impurity with respect to that of the fuel and of the poloidal stagnation point of the flow velocity.
For the ITER simulation, where the SOL heat flux channel width is larger (fixed by the cross-field transport assumptions in the model), ionization of fuel and impurity atoms occurs primarily in the divertor SOL, close to the target, keeping T e high further upstream. This in turn means that the different impurity (Wm -3 ) Fig. 9. Distributions of total radiation in the divertor for Ne (left) and N (right) seeding at P SOL = 100 MW, high p n ∼10 Pa and similar P rad,div , c cimp (see Fig. 8  for the location of the cases in the main database). Whilst most of the radiation is localized to the divertor, the higher spread in the Ne case is clear. charge states exist closer to the targets in ITER and the radiation is better confined at both inner and outer targets. Further simulations are planned to study the consequences of reduced λ q on this screening in ITER.
Momentum losses Section 3. 1. 3 discussed the dependence of the target heat loading with varying fuel throughput and impurity seeding, showing the classical evolution of partial detachment for vertical target configurations with increasing levels of dissipation. Fig. 7  illustrated the progressive reduction in p tot,t /p tot,u with increasing p n for one specific value of c Ne in the Ne seeded database. Fig. 10  now extends this to the entire Ne simulation set, but now plotting the pressure ratio on each flux surface of the numerical grid outside the separatrix within the first 2 cm of the omp SOL as a function of T e,t on that flux surface at both the IT and OT, excluding the first SOL ring outside the separatrix (at (r-r sep ) omp = 1. 2 mm) where radial diffusion into the PFR is strongest.
05 at T e,t ∼0. 5 eV, the lowest values achieved in the simulations and the range of T e,t over which volume recombination is expected to play a role in ion removal. The arbitrarily chosen analytic form: = ( ) f A e 1 1 mom loss T T n / * e t , with coefficients A, T* and n, is shown in  [69]  to reproduce well the shape of the pressure loss fraction vs. T e,t curves for the range of modelling and experimental studies considered there. These fits (modelling cases only), reproduced from Fig. 15  in  [69]  have been included on the ITER simulation data in Fig. 10  for comparison, along with a best (non-linear least squares) fit of the above expression to the ITER points across the whole span of T e,t (for which A = 0. 87, T* = 0. 43, n = 5. 30 and the goodness of fit parameter R 2 = 0. 96). Evidently, the ITER dataset can also be well represented by this trial function.
Fig. 11  shows that a similar trend is found in the ITER simulation database, over the same radial extent of the SOL as for the momentum losses in Fig. 10  (1 ≤ (r-r sep ) omp ≤ 18 mm) and for both targets. A fit of the function = T Bn e t D t , 2, to all data with T e,t ≤ 10 eV yields B = 4. 97 × 10 13 and β = -0. 66 for R 2 = 0. 93, thus similar to the coefficients β found for a large variety of code studies in  [69, 81]. A fit of the same quality is obtained whether or not the first 2-3 flux tubes outside the separatrix are included (where diffusion into the PFR is strongest), but if all points above T e,t = 10 eV are also used, the data deviate very significantly from the fit. A potentially important caveat here, however, is that the SOLPS-4. 3 simulations have been performed assuming a Be coated divertor target (see Section 3.
1. 1), which has ∼6x. This curve may be compared with those in Fig. 16  of  [69]. lower energy reflection coefficient for atomic deuterium than W  [84]  and behaves very similarly to carbon, for which the code results discussed in  [69, 81]  have been obtained. In this context, a study has been performed  [54]  with SOLPS-ITER in which, for two example cases in the SOLPS-4. 3 database (at low and high p n ), a fully Be-coated divertor is compared with the other extreme, a pure W target (but in neither case is target sputtering switched on). The principal differences appear indeed to be in the ratio of molecules to atoms in the target recycling flux (higher for Be), determined by the reflection and recombination properties of the different materials, which increases the molecular density in the divertor region for Be surfaces compared with W.
0 simulation dataset from the simulations of a series of slot divertor configurations, found as strong a link between volumetric power losses and T e,t as that between T e,t and n D2,t. The fact that most of the power dissipation in the ITER simulations proceeds through seeded impurity radiation rather than from the deuterium atomic and molecular radiation (see e. g. Fig. 9 ), supports volumetric power losses as a key physical process required to obtain low T e,t. As for VR, which becomes very significant for T e ≲ 1 eV and which therefore might be expected to be playing a major role in the rapid decrease of 1 -f mom-loss at low T e,t (Fig. 10 ), new "balance analysis" is underway in which the different contributions to particle and energy losses in any given flux tube in the code results are examined in detail.
This capability, recently introduced into the SOLPS-ITER code version, can only currently be performed on output from that version so that new cases, equivalent to the SOLPS-4. 3 runs, need to be generated. A series of such cases covering the range of p n and c Ne encompassed by the SOLPS-4. 3 database is in preparation, using the same physics model in SOLPS-ITER as that used in SOLPS-4. 3 to generate the main database. Preliminary analysis of the available converged new runs at high p n indicates that the VR is more important at the inner divertor than the outer; for example, in a simulation for p n ∼ 10 Pa, integrating along the 3rd SOL flux ring outside the separatrix (corresponding to q ⊥, pk at the outer target), the VR sink in the outer divertor is ∼20% of the outer divertor particle source, while the VR sink in the inner divertor is ∼80% of the inner source.
Further analysis of more simulation runs is required, and is planned, including the application of the 2PMF approach proposed in  [69]  and a sensitivity scan where VR is turned off. Ultimately, the hope is that reduced models, along the lines advocated in  [69] , might be used to generate solutions for ITER, based on SOLPS-4. 3 and forthcoming SOLPS-ITER simulations. This would naturally provide a much more computationally rapid framework suitable for inclusion in integrated models of the ITER plasma which require simpler descriptions of the divertor response than obtained with full plasma boundary code runs, but which must capture the essential (at least two-dimensional) physics at play. Such reduced models need data such as those shown in Figs. 10  and 11. This of course represents a different approach to the scalings which have been developed on the basis of the global simulation database for use as input to core modelling and which have constituted the mainstay of integrated modelling for the ITER divertor thus far.
Narrow heat flux channels and drifts Considering still the case of an unperturbed magnetic equilibrium, there are two further factors which may influence the target heat loading in addition to component shaping: heat flux channel width and fluid drifts. Beginning with the former, Section 3. 1. 1 has discussed how empirical scaling on current devices and ion magnetic drift theory for the upstream heat flux width extrapolate to λ q ∼ 1 mm for the ITER burning plasma baseline at I p = 15 MA. Turbulence simulations using the XGC1 and BOUT++ codes find, in contrast, that the ion orbit prediction is broken at the ITER scale and that λ q ∼ 5-6 mm might be expected, closer to the λ q ∼ 3. 5 mm fixed in the SOLPS-4. 3 database (Fig. 4 ). To examine the consequences for power handling of a much lower λ q , cross-field transport coefficients can be reduced in the ITER SOLPS modelling.
This numerically challenging exercise is being attempted at the time of writing with the new SOLPS-ITER code, for the W/Be environment, but the results are insufficiently mature to be presented here. Instead, recourse can be made to an earlier SOLPS-4. 3 study (thus no drifts) made for the case of carbon divertor plates and P SOL = 100 MW  [28]. In that work, fuelling throughput scans were performed on a grid with finer resolution close to the separatrix, and with D ⊥ and χ ⊥ reduced to 50% and 25% of the standard values (D ⊥ = 0. 3 m 2 s -1 , χ ⊥ = 1. 0 m 2 s -1 , Section 3. 1. 1). A subset of the results reported in  [28]  have been extracted for comparison with the SOLPS-4. 3 "carbon-free" database, focussing on q pk shp , , and are shown in Fig. 14 , which is a replot of Fig. 13  (note the extended y-axis) and which also includes some new points obtained from SOLPS-ITER drift runs (see below).
3 database at P SOL = 100 MW. The 8 points available at the time of writing are included in Fig. 14 , discussed above in relation to the consequences of reduced transport (without drifts). The effect of target shaping has been fully accounted for. A detailed description of these code runs, deeper physics analysis and a discussion of the differences between these new simulations and their non-drift SOLPS-ITER and SOLPS-4. 3 counterparts may be found in  [90]. Apart from the activation of drifts, the new SOLPS-ITER simulations differ in an important aspect from those obtained with SOLPS-4. 3: they include an H-mode transport barrier inside the separatrix, obtained by specifying radial profiles of D ⊥ and χ ⊥ tuned to provide approximately the expected Q DT = 10 pedestal profiles of n e , T e and T i.
Inclusion of the transport barrier is important given the impact on drifts of the strong electric potential gradients which develop in the H-mode. In the SOL, D ⊥ = 0. 3 m 2 s -1 , χ ⊥ = 1. 0 m 2 s -1 are fixed, as in the SOLPS-4. 3 runs, producing a similar near SOL omp λ q ∼ 4 mm. Unlike the SOLPS-4. 3 simulations, the SOLPS-ITER runs assume W at the divertor targets (see discussion in Section 3. 1. 1), though W impurity is not evolved. They also use Ne impurity seeding, but it has not yet been possible to produce a well-defined series of throughput scans at varying c Ne as in the SOLPS-4. 3 database. One or two equivalent N-seeded drift runs are available but are not included in Fig. 14  (see Section 3. 1. 4 and  [78] ). The trend in the new runs is similar to the non-drift cases: with increasing p n , the out-in asymmetry reduces.
As shown in  [78] , the divertor radiation distributions with and without drifts are very similar at the higher p n for approximately equivalent c Ne. What is clear is that if these new results can be trusted, they are a further indication that operation at higher p n is likely to be required at ITER from the point of view of power handling. Drift runs at low SOL transport have not yet been attempted, but are planned and are clearly a necessary step. An important question to address is the extent to which, in pushing to higher p n and hence deeper detachment than covered in the existing ITER SOLPS-4. 3 database, the radiation distribution moves towards the X-point, eventually forming an X-point MARFE, as seen commonly on current devices (see e. g. [73, 74] ) at high seeding rates. Here, apart from the impact of increased radiation inside the confinement region/pedestal, which may prevent strongly detached regimes on ITER given the proximity to the L-H power transition threshold at which the device will operate, a key issue is the stability of the detachment front.
From the perspective of detachment control, it will be important to assess whether or not the sharp transitions/bifurcations ("detachment cliff") which have been observed experimentally in some current devices, may also appear on ITER at higher levels of dissipation than considered in the current database. In fact, recent modelling of the bifurcations seen on the DIII-D tokamak during H-mode operation with the UEDGE code can only qualitatively match experiment if drifts and currents are activated in the simulations  [91]. As mentioned earlier, the studies reported in  [28]  with SOLPS-4. 3 at reduced transport used a more refined grid in the separatrix region. This issue of grid resolution is a further factor which can influence q pk sym , , even in the case of "normal" transport assumptions.
A recent study of the impact of grid discretization error using SOLPS-4. 3 found a ∼25% increase in q pk sym , when the grid was refined by a factor 4 for partially detached, low power, D-only ITER "test" cases  [92]. Further tests with SOLPS-ITER are underway at the time of writing for the full power, impurity seeded cases. Recent work using a large database of measurements from the JET and ASDEX Upgrade tokamaks finds clear evidence, supported by quantitative arguments, for a central role of the upstream separatrix density normalized to the Greenwald density, n e,sep /n GW , in determining the onset of the H-L back transition  [93]. This H-mode density limit is strongly correlated with the approach to the ideal ballooning instability threshold at the plasma edge. Fig. 15a  gives the dependence of n e,sep (evaluated at the omp) on p n for the simulation cases compiled in Fig.
[74] ). This is most likely related to the difficulty of measuring n e,sep itself under the high power H-mode conditions in which most seeding work is performed on current devices. The strong dependence of c imp (=c Ne ) on n e,sep seen in Fig. 15a  is examined further in Fig. 15b  for the Ne-seeded simulations only, since this represents the largest dataset in which c imp is systematically varied. Here the data are grouped according to their value of P rad,div within 5 bands of width ∼7 MW and centred on 30, 37, 44, 52 and 59 MW, spanning the whole range of P rad,div (from 26. 3 → 62. 5 MW). A fit to the function c Ne = An e,sep has been performed for the data in each of the 5 P rad,div intervals. Values of α range from -2. 0 → -4. 0, with the n e,sep 2 dependence found for the highest P rad,div where detachment is strongest (i.
The analysis in  [93] , finds the observed H-mode operational boundary to be roughly limited by n e,sep /n GW < 0. 4-0. 5 in JET and ASDEX Upgrade, when the MHD ballooning parameter at the separatrix location reaches values in the range of theoretical predictions for ballooning instability. This observed critical n e,sep /n GW can be quantitatively derived using the experimental λ q scaling  [56] , itself obtained partly from JET and ASDEX Upgrade data. If the same limit also applies to ITER, then the lowest seeding values in the SOLPS-4. 3 database would be marginal, though these were obtained for SOL transport assumptions giving higher λ q than the empirical scaling. At reduced transport (the carbon case at f perp = 0. 25 in Fig. 15a ), giving λ q = 1. 3 mm, closer to the empirical scaling prediction for ITER at 15 MA, n e,sep exceeds the predicted limit and H-mode operation may be compromised.
Fixing a budget in terms of bulk recrystallization may also have the possible added benefit of raising the threshold for surface cracking due to the fast, repetitive transient heat loads brought to the surface by ELMs (see Section 5) since the cracking itself is facilitated by the reduced mechanical strength of recrystallized material. Fig. 17  makes the transformation of q pk shp , vs. p n from the SOLPS-4. 3 database to T surf vs. p n , using the data from Fig. 13 , thus taking into account the impact of MB surface shaping and target tilting. Here T surf is the surface temperature at the centre of the MB, where the macrocracks usually form, and is thus the appropriate location at which to apply the recrystallization criterion. The transformation from q pk shp , to T surf is performed using a simple linear relationship (T surf (°C) = 85.
These qualification tests typically subject the MBs to several hundred such cycles in comparison with the several thousand cycles for the baseline load. Physically, this is equivalent to moving to lower divertor neutral pressure in the p n vs. q pk , operating domain in Figs. 5a, 13 and 14. Such events may occur on relatively short timescales (hundreds of ms to seconds), but are not to be confused with the fast (ms) transients, ELMs and disruptions, discussed in the following section. As discussed in Section 3. 2. 1, lower p n corresponds to deposited power on the MB surface dominated by the thermal plasma and ion recombination components of the total heat load (Fig. 12 ), both sensitive to divertor target tilting and MB top surface shaping. At the lowest p n (∼2 Pa) in the SOLPS-4. 3 database (with λ q ∼3.
In addition, the MB surface shaping, then only a proposal, has now been extensively studied and fixed into the design (see Section 3. 2. 1). On the physics side, there have nevertheless been significant developments since 2013, particularly with regard both to an improved understanding of the expected Type I ELM target energy fluxes and to their impact on the W material. For the former, an important advance has been the publication of a new multi-machine (ASDEX Upgrade, JET and MAST) scaling  [10]   with n e,ped [x10 20 m -3 ], T e,ped [keV] the electron density and temperature at the H-mode pedestal top, ΔW ELM the ELM energy loss (in %) relative to the plasma stored energy (ΔW ELM = W ELM /W plasma ) and R the major radius  [m]. The scatter of this data about the regression is roughly a factor of 3, for a range in ε || of more than two orders of magnitude in the database.
The data is yet relatively sparse, but observations on ASDEX Upgrade  [10] , DIII-D  [131]  and JET  [132]  generally support the scaling applicability also at the inner divertor target. On ASDEX Upgrade, an in-out symmetry of ∼1. 25 is observed in the peak target ELM energy fluences, consistent with approximate symmetry in the incoming ε || (since the projected area of the inner and outer target flux tubes differs by the inverse ratio of the toroidal fields at the inboard and outboard targets). The total ELM energy deposited on the IT in ASDEX Upgrade is about a factor of 2 higher than at the outboard since the ELM footprint mapped to the omp is twice as high as that for the OT (hence a broader inner target deposition profile)  [132]. On ASDEX Upgrade, the scaling is equally found to apply to ELMs mitigated by magnetic perturbations, though the database is again still relatively limited  [10].
In the IRP, the latter is now estimated to be around 6 years of fusion power operation beyond the initial non-active phase campaigns. This corresponds to ∼2000 h of projected high performance plasmas, during which the divertor will need to function in its intended, partially detached operating mode, with extrinsic impurity seeding used as the principal mechanism for volumetric power dissipation to maintain acceptable target stationary heat flux densities. Making use of the database of SOLPS-4. 3 simulations compiled for "carbon-free" Ne and N-seeded operation under nominal burning plasma conditions (P SOL = 100 MW)  [3] , the paper has focused on the accessible domain in terms of two key operational parameters, divertor neutral pressure and peak target heat flux density, each tightly linked to the other.
This paper has tried to reflect some of this advance whilst at the same time indicating where more research is required. To conclude the article, some of these key areas are listed below in abbreviated form, beginning with plasma physics issues and ending with a point on outstanding (W) material aspects, also raised explicitly in the review of  [125] : • Improve understanding of the expected near-SOL heat flux width at high plasma current in ITER; • Extend database of ITER divertor simulations with fluid drifts, in- cluding cases with reduced transport to examine consequences of very narrow λ q ; • Continue numerical and experimental studies to confirm that N and Ne will be equally effective divertor radiators for ITER at high performance, but only within ∼factor 2 of the H-L transition threshold power.
Acknowledgements The views and opinions expressed herein do not necessarily reflect those of the  ITER Organization. ITER is the nuclear facility INB 174. Portions of this work were performed under the auspices of the  U. S. Department of Energy  under grants  DE-SC0012315  and  DE-SC0013911 , and by funding of the  Department of Engineering Physics and of the College of Engineering  at the  University of Wisconsin-Madison, USA. The work of the SPbPU group was supported by the  Russian Science Foundation , grant no. 17-12-01020. Several of the co-authors have contributed within the framework of the  ITER Scientist Fellowship Network  (ISFN). The authors would like to specifically acknowledge the very significant contribution to the ITER divertor design activity made over more than 2 decades by H.
1993) , and molecular phylogenetic approaches  (Holford et al. 2009; Puillandre et al. 2011; Castelin et al. 2012). A striking diversity of anterior digestive system anatomies among Terebridae, has led  Rudman (1969)  to question the monophyly of the family, and establish the family Pervicaciidae mainly for species of the genus Duplicaria lacking the venom gland, and possessing an ancestral radula type. The distinctive features of Pervicaciidae were further acknowledged by  Taylor (1990) , however, its independent origin from the rest of Terebridae was rejected based on an extensive morphological dataset  (Taylor et al. 1993). The monophyly of Terebridae, including Duplicaria was further confirmed by molecular studies  (Holford et. al. 2009; Castelin et al. 2012) , which recognized Duplicaria as one of the early diverging lineages of the family, together with an enigmatic deep-water species  'Terebra' jungi Lai, 2001 (Holford et al.