Development of a radiative divertor for DIII-D

Scrapeoff Layer (SOL) data from DIII-D and C-Mod have been acquired and analyzed for radial particle transport based on a particle balance model. This has allowed a detailed comparison for L-mode plasmas. The inferred radial particle flux, Γ ⊥ (r), is parameterized in terms of diffusive [D eff (r) ≡ Γ ⊥ (r)/∇n(r)]and convective particle transport [v eff (r) ≡ Γ ⊥ (r)/n(r)]. The magnitude of the inferred D eff or v eff increases across the SOL for both tokamaks. The inferred D eff or v eff in the 'far' SOL (one density e-folding length from the separatrix and beyond) are essentially unchanged by changes in core density by factors of 2-3. This corresponds to changes in the far SOL density, collisionality (ν*), and radial fluxes of a factor of 10 or more. Thus ν* does not appear to be an important parameter in determining the radial particle transport in that region.The dimensionlessly-scaled SOL plasma profiles from the two tokamaks overlay for similar dimensionless plasma parameters. The SOL density profile near the separatrix is steeper than in the 'far' SOL. The scaled D eff and v eff are slightly larger on DIII-D than C-Mod. This difference appears to be within experimental uncertainties.Neutral ionization in the SOL does not appear to affect radial transport but may be related to the observed flattening of the density profiles with increasing n e .

show abstract

“…Whyte [14], also indicate R leak is small for DIII- lower divertor design and modeling which indicates that divertor leakage is small [30,31].…”

Section: Experimental Input To the Value Of F Cmentioning

confidence: 95%

Comparison of particle transport in the scrape-off layer plasmas of Alcator C-Mod and DIII-D

Lipschultz

Whyte

LaBombard

2005

Plasma Phys. Control. Fusion

101

View full text Add to dashboard Cite

show abstract

“…(We use 'core' to refer to all the plasma inside the separatrix.) The leakage of impurity gas into the core plasma should be reduced in DIII-D with the completion of the Radiative Divertor Project (RDP) [9], which will include a closed divertor to trap gas in the divertor legs. The upper outer baffle and cryopump of the RDP have been installed.…”

Section: Overviewmentioning

confidence: 99%

Survey of target plate heat flux in diverted DIII-D tokamak discharges

et al. 1998

View full text Add to dashboard Cite

A series of observations is presented concerning divertor heat flux, q div , in the DIII-D tokamak, and it is shown that many features can be accounted for by assuming that the heat flux flows preferentially along field lines because τ < τ ⊥ in the scrape-off layer (SOL). Exceptions to this agreement are pointed out and the discrepancies explained by means of two dimensional (2-D) effects. About 80% of the discharge input power can be accounted for. The power deposited on the target plate due to enhanced losses during edge localized modes (ELMs) is less than 10% of the total target power in most cases. X point height scans for lower single null (LSN) diverted discharges show that the peak heat flux variation is primarily due to flux expansion and secondarily due to transport of energy across the magnetic field in the divertor. At the outer strike point q div,peak ∝ Pin(Ip − Ip,0)G(gin)(1/Bt) 4/9 (B div /Bmp)f (L div , χ ⊥ ), where G is a linear function of the inner gap, gin, over a specified range and f describes cross-field energy transport in the divertor. Evidence of radial in-out asymmetries (comparing the outer strike point with the inner strike point or centre-post) and toroidal asymmetries in q div is presented and the heat flux peaking due to tile gaps and misalignment of tiles is examined. For magnetically balanced double null (DN) discharges with downward ∇B ion drift, it is found that q div is inherently higher in the lower divertor than in the upper divertor, having a 3:1 downward bias. Examples of heat flux reduction by gas puffing deuterium or neon in LSN and DN discharges are given. At least a threefold reduction of the peak heat flux in both the upper and lower divertors of a DN discharge, using D2 puffing, is reported.

show abstract

“…It was desirable to enhance the divertor design to permit control of the recycling neutrals, and thus better control of the neutral density fueling the core plasma. The UEDGE code was used to assist the design of this "Advanced Divertor Project" [47,48]. Code simulations indicated the neutral leakage from the divertor could be controlled with a baffle around the outer strike point.…”

Section: D Model (Uedge)mentioning

confidence: 99%

Developing Models for the DIII-D Boundary Plasma

Porter

Rognlien

Rensink

et al. 2005

Fusion Science and Technology

View full text Add to dashboard Cite

Abstract. Development of the comprehensive codes used to study the boundary region of the DIII-D tokamak has been done in parallel with improvement of the diagnostics of this important region of the plasma. These codes have been used to interpret the diagnostic data and assist in design of improved divertor configurations. The development of codes used for analysis on DIII-D is described briefly. Model validation by comparing with the extensive DIII-D boundary region diagnostic data is also discussed.

show abstract

Development of a radiative divertor for DIII-D

Cited by 32 publications

References 3 publications

Comparison of particle transport in the scrape-off layer plasmas of Alcator C-Mod and DIII-D

Comparison of particle transport in the scrape-off layer plasmas of Alcator C-Mod and DIII-D

Survey of target plate heat flux in diverted DIII-D tokamak discharges

Developing Models for the DIII-D Boundary Plasma

Contact Info

Product

Resources

About