Physics based integrated modelling of the baseline scenario for a Fusion Nuclear Science Facility based on the Advanced Tokamak concept (FNSF-AT) (Chan et al 2010 Fusion Sci. Technol. 57 66) has ...found steady-state equilibria with good stability and controllability properties at the fusion performance required to accomplish FNSF's nuclear science mission with margin. 2D divertor analysis for this baseline scenario predicts that peak heat flux <10 MW m−2 can be obtained even with scrape-off layer power width ∼1 mm. Using this baseline fusion performance, high fidelity and high-resolution 3D neutronics calculations show acceptable cumulative end-of-life organic insulator dose levels in all the device coils, and TBR >1. Two current drive scenarios, two divertor configurations, and two blanket concepts have been analysed. FNSF-AT would complement ITER in addressing science and technology gaps to a commercially attractive DEMO, and could enable a DEMO construction decision triggered by the achievement of Q = 10 in ITER.
Abstract
Edge codes such as SOLPS coupled to neutral codes such as EIRENE have become so comprehensive and sophisticated that they now constitute, in effect, ‘code-experiments’ that, as for actual ...experiments, can benefit from interpretation using simple models and conceptual frameworks, i.e. reduced models. The first task is the identification of options for the reduced model control parameters that are best suited for control of the action of the divertor, i.e. for control of target power loading and sputter-erosion, primarily. A strong correlation between the electron temperature at the divertor target,
T
e,t
, and the neutral deuterium D
2
density at the target,
n
D2,t
, flux-tube resolved, has recently been reported for a number of code studies including SOLPS-4.3 modeling of a set of ∼50 ITER baseline cases:
Q
DT
= 10,
q
95
= 3,
P
SOL
= 100 MW, metallic walls, and Ne seeding (Pitts
et al
2019
Nucl. Mater. Energy
20
100696). This part A of the present study reports new results for largely the same ITER cases, confirming the strong correlations reported earlier between local values of
T
e,t
, and (i)
n
D2,t
, and (ii) normalized volumetric losses of power and pressure in the divertor. Strong correlations have now also been found, and are reported here for the first time, between
T
e,t
and
all
of the divertor target quantities of practical interest. A physical explanation for this surprising result has not been fully identified; nevertheless it has encouraging implications for reduced modeling of the ITER divertor. For such ITER conditions, (i) the global Ne injection rate, Inj
Ne
(Ne s
−1
), and (ii) the electron temperature at the location on the target where the peak power deposition occurs,
T
e,t
@q
⊥,pk
(eV), are found to be promising reduced model control parameters. In the companion report, part B, a reduced model for the ITER divertor is developed and described in detail, based on reversed-direction 2 point modelling, Rev2PM. The input to the reduced model is a value of the variable pair
T
e
,
t
@
q
⊥
,
p
k
,
I
n
j
Ne
and the output are values of the various target as well as divertor-entrance quantities of practical interest, e.g.
q
⊥,pk
,
n
e,Xpt
(the electron density at the poloidal location of the X-point), etc. In part B the reduced model is quantitatively characterized using one half of the code cases; it is then used to successfully predict (replicate) the code values of e.g.
n
e,Xpt
for the other half of the cases.
Abstract
Edge codes such as SOLPS coupled to neutral codes such as EIRENE have become so comprehensive and sophisticated that they now constitute, in effect, ‘code-experiments’ that, as for actual ...experiments, can benefit from interpretation using simple models and conceptual frameworks, i.e. reduced models. The first task is the identification of options for the reduced model control parameters that are best suited for control of the action of the divertor, i.e. for control of target power loading and sputter–erosion, primarily. A strong correlation between the electron temperature at the divertor target,
T
e,t
, and the neutral deuterium D
2
density at the target,
n
D2,t
, flux-tube resolved, has recently been reported for a number of code studies including SOLPS-4.3 modeling of a set of ∼50 ITER baseline cases:
Q
DT
= 10,
q
95
= 3,
P
SOL
= 100 MW, metallic walls, and Ne seeding (Pitts
et al
2019). Part A of the present study reports new results for largely the same ITER cases, confirming the strong correlation reported earlier between local values of
T
e,t
, and (i)
n
D2,t
, and (ii) normalized volumetric losses of power and pressure in the divertor. Strong correlations have now also been found, and are reported here for the first time, between
T
e,t
and
all
of the divertor target quantities of practical interest. A physical explanation for this surprising result has not as yet been fully identified; nevertheless it has encouraging implications for reduced modeling of the ITER divertor. For such ITER conditions, (i) the global Ne injection rate, Inj
Ne
(Ne s
−1
), and (ii) the electron temperature at the location on the target where the peak power deposition occurs,
T
e,t
@q
⊥,pk
(eV), are found to be promising reduced model control parameters. In this part B, a reduced model for the ITER divertor is developed and described in detail, based on reversed-direction two point modeling, Rev2PM. The input to the reduced model is a value of the variable pair
T
e
,
t
@
q
⊥
,
p
k
,
I
n
j
Ne
for a chosen case and the output are values of the various target as well as divertor-entrance quantities of practical interest, e.g.
q
⊥,pk
, the electron density at the X-point,
n
e,Xpt
, etc. The reduced model was quantitatively characterized using one half of the code cases; it was then used to successfully predict (replicate) the code values of e.g.
n
e,Xpt
for the other half.
Linear plasma generators are cost effective facilities to simulate divertor plasma conditions of present and future fusion reactors. They are used to address important R&D gaps in the science of ...plasma material interactions and towards viable plasma facing components for fusion reactors. Next generation plasma generators have to be able to access the plasma conditions expected on the divertor targets in ITER and future devices. The steady-state linear plasma device MPEX will address this regime with electron temperatures of 1-10 eV and electron densities of 1021​−​1020 m−3. The resulting heat fluxes are about 10 MW m−2. MPEX is designed to deliver those plasma conditions with a novel Radio Frequency plasma source able to produce high density plasmas and heat electron and ions separately with electron Bernstein wave (EBW) heating and ion cyclotron resonance heating with a total installed power of 800 kW. The linear device Proto-MPEX, forerunner of MPEX consisting of 12 water-cooled copper coils, has been operational since May 2014. Its helicon antenna (100 kW, 13.56 MHz) and EC heating systems (200 kW, 28 GHz) have been commissioned and 14 MW m−2 was delivered on target. Furthermore, electron temperatures of about 20 eV have been achieved in combined helicon and ECH heating schemes at low electron densities. Overdense heating with EBW was achieved at low heating powers. The operational space of the density production by the helicon antenna was pushed up to 1.1×1020 m−3 at high magnetic fields of 1.0 T at the target. The experimental results from Proto-MPEX will be used for code validation to enable predictions of the source and heating performance for MPEX. MPEX, in its last phase, will be capable to expose neutron-irradiated samples. In this concept, targets will be irradiated in ORNL's High Flux Isotope Reactor and then subsequently exposed to fusion reactor relevant plasmas in MPEX.
Divertor design and choice of plasma-facing materials (PFM) will be essential to the success of next-generation fusion reactors as they operate under more powerful scenarios. Understanding and ...controlling interactions between the plasma and PFM is essential to making these choices. Within these plasma–material interactions and especially in tungsten (W), the interplay between the most abundant plasma species (hydrogen isotopes and helium, He) with the wall material alters fuel retention. However, this interplay is yet to be sufficiently understood to confidently project fuel retention levels to future fusion devices. The paper presents a series of integrated simulations of fusion plasmas and their interaction with tungsten. Specifically, this study assesses the impact of He plasma pre-exposure on hydrogenic species retention during 100 s of burning plasma operations (BPO) in ITER. Multiple pre-exposure scenarios are considered, including sub-surface damage resulting from exposures in the linear device PISCES and from early ITER He-operation. The predictions from these consecutive He-BPO exposures show that fuel content and spatial distribution in the material are largely determined by the He-induced damage, as manifest in: (i) changes in surface temperature expected during BPO have little effect on fuel retention in the presence of He-induced damage; (ii) gas content stabilizes quickly in substrates pre-exposed in PISCES, at levels set by the concentration of pre-existing vacancies, while it continues to increase in substrates initially pristine or pre-exposed to ITER He plasmas; (iii) the presence of He and He–V clusters in the near-surface region locally increases hydrogenic retention, but decreases its permeation; this results in hydrogenic species that remain closer to the surface in pre-damaged substrates, while the bulk content is higher for initially pristine cases. In summary, the interaction and binding of D and T with the pre-existing He–V clusters modifies retention and permeation of hydrogen species during ITER BPO.
In this paper, the effects of a pre-discharge lithium evaporation variation on highly shaped discharges in the National Spherical Torus Experiment (NSTX) are documented. Lithium wall conditioning ...(‘dose’) was routinely applied onto graphite plasma facing components between discharges in NSTX, partly to reduce recycling. Reduced Dα emission from the lower and upper divertor and center stack was observed, as well as reduced midplane neutral pressure; the magnitude of reduction increased with the pre-discharge lithium dose. Improved energy confinement, both raw τE and H-factor normalized to scalings, with increasing lithium dose was also observed. At the highest doses, we also observed elimination of edge-localized modes. The midplane edge plasma profiles were dramatically altered, comparable to lithium dose scans at lower shaping, where the strike point was farther from the lithium deposition centroid. This indicates that the benefits of lithium conditioning should apply to the highly shaped plasmas planned in NSTX-U.
A sequence of H-mode discharges with increasing levels of pre-discharge lithium evaporation (‘dose’) was conducted in high triangularity and elongation boundary shape in NSTX. Energy confinement ...increased, and recycling decreased with increasing lithium dose, similar to a previous lithium dose scan in medium triangularity and elongation plasmas. Data-constrained SOLPS interpretive modeling quantified the edge transport change: the electron particle diffusivity decreased by 10–30x. The electron thermal diffusivity decreased by 4x just inside the top of the pedestal, but increased by up to 5x very near the separatrix. These results provide a baseline expectation for lithium benefits in NSTX-U, which is optimized for a boundary shape similar to the one in this experiment.
Pedestal fueling through edge recycling is examined with the interpretive OEDGE code for high-density discharges in DIII-D. A high current, high-density discharge is found to have a similar radial ...ion flux profile through the pedestal to a lower current, lower density discharge. The higher density discharge, however, has a greater density gradient indicating a pedestal particle diffusion coefficient that scales near linear with 1/Ip. The time dependence of density profile is taken into account in the analysis of a discharge with low frequency ELMs. The time-dependent analysis indicates that the inferred neutral ionization source is inadequate to account for the increase in the density profile between ELMs, implying an inward density convection, or density pinch, near the top of the pedestal.
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