TOKES studies of the thermal quench heat load reduction in mitigated ITER disruptions

Disruption mitigation by massive gas injection (MGI) of Ne gas has been simulated using the 3D TOKES code that includes the injectors of the Disruption Mitigation System (DMS) as it will be implemented in ITER. The simulations have been done using a quasi-3D approach, which gives an upper limit for the radiation heat load (notwithstanding possible asymmetries in radial heat flux associated with MHD). The heating of the first wall from the radiation flash has been assessed with respect to injection quantity, the number of injectors, and their location for an H-mode ITER discharge with 280 MJ of thermal energy. Simulations for the maximum quantity of Ne (8 kPa m3) have shown that wall melting can be avoided by using solely the three injectors in the upper ports, whereas shallow melting occurred when the midplane injector had been added. With all four injectors, melting had been avoided for a smaller neon quantity of 250 Pa m3 that provides still a sufficient radiation level for thermal load mitigation

An updated analysis of NN elastic scattering data to 1.6 GeV

An energy-dependent and set of single-energy partial-wave analyses of $NN$ elastic scattering data have been completed. The fit to 1.6~GeV has been supplemented with a low-energy analysis to 400 MeV. Using the low-energy fit, we study the sensitivity of our analysis to the choice of $\pi NN$ coupling constant. We also comment on the possibility of fitting $np$ data alone. These results are compared with those found in the recent Nijmegen analyses. (Figures may be obtained from the authors upon request.)Comment: 17 pages of text, VPI-CAPS-7/

Multi-machine scaling of the main SOL parallel heat flux width in tokamak limiter plasmas

As in many of today’s tokamaks, plasma start-up in ITER will be performed in limiter configuration on either the inner or outer midplane first wall (FW). The massive, beryllium armored ITER FW panels are toroidally shaped to protect panel-to-panel misalignments, increasing the deposited power flux density compared with a purely cylindrical surface. The chosen shaping should thus be optimized for a given radial profile of parallel heat flux, q in the scrape-off layer (SOL) to ensure optimal power spreading. For plasmas limited on the outer wall in tokamaks, this profile is commonly observed to decay exponentially as q q = − exp ( / r λ ) 0 q omp , or, for inner wall limiter plasmas with the double exponential decay comprising a sharp near-SOL feature and a broader main SOL width, λq omp. The initial choice of λq omp , which is critical in ensuring that current ramp-up or down will be possible as planned in the ITER scenario design, was made on the basis of an extremely restricted L-mode divertor dataset, using infra-red thermography measurements on the outer divertor target to extrapolate to a heat flux width at the main plasma midplane. This unsatisfactory situation has now been significantly improved by a dedicated multi-machine ohmic and L-mode limiter plasma study, conducted under the auspices of the International Tokamak Physics Activity, involving 11 tokamaks covering a wide parameter range with R = = 0.4–2.8 m, 1 B I 0 p .2–7.5 T, = 9–2500 kA. Measurements of λq omp in the database are made exclusively on all devices using a variety of fast reciprocating Langmuir probes entering the plasma at a variety of poloidal locations, but with the majority being on the low field side. Statistical analysis of the database reveals nine reasonable engineering and dimensionless scalings. All yield, however, similar predicted values of λq omp mapped to the outside midplane. The engineering scaling with the highest statistical significance, λ = ( / ( )) ( / /κ) − − q 10 P V W m a R omp tot 3 0.38 1.3 , dependent on input power density, aspect ratio and elongation, yields λq omp = [7, 4, 5] cm for Ip = [2.5, 5.0, 7.5] MA, the three reference limiter plasma currents specified in the ITER heat and nuclear load specifications. Mapped to the inboard midplane, the worst case (7.5 MA) corresponds to λq ~ 57 1 ± 4 imp mm, thus consolidating the 50mm width used to optimize the FW panel toroidal shape.EURATOM 633053Czech Science Foundation GA CR P205/12/2327, GA15-10723S, MSMT LM2011021US Department of Energy DE-FG02- 07ER54917, DE-AC02-09CH11466, DE-FC02-04ER5469

Main chamber wall plasma loads in JET-ITER-like wall at high radiated fraction

Future tokamak reactors of conventional design will require high levels of exhaust power dissipation (more than 90% of the input power) if power densities at the divertor targets are to remain compatible with active cooling. Impurity seeded H-mode discharges in JET-ITER-like Wall (ILW) have reached a max- imum radiative fraction ( F rad ) of ∼75%. Divertor Langmuir probe (LP) measurements in these discharges indicate, however, that less than ∼3% of the thermal plasma power reaches the targets, suggesting a missing channel for power loss. This paper presents experimental evidence from limiter LP for enhanced cross-field particle fluxes on the main chamber walls at high F rad . In H-mode nitrogen-seeded discharges with F rad increasing from ∼30% to up to ∼75%, the main chamber wall particle fluence rises by a factor ∼3 while the divertor plasma fluence drops by one order of magnitude. Contribution of main chamber wall particle losses to detachment, as suggested by EDGE2D-EIRENE modeling, is not sufficient to explain the magnitude of the observed divertor fluence reduction. An intermediate detached case obtained at F rad ∼60% with neon seeding is also presented. Heat loads were measured using the main chamber wall thermocouples. Comparison between thermocouple and bolometry measurements shows that the frac- tion of the input power transported to the main chamber wall remains below ∼5%, whatever the divertor detachment state is. Main chamber sputtering of beryllium by deuterium is reduced in detached condi- tions only on the low field side. If the fraction of power exhaust dissipated to the main chamber wall by cross-field transport in future reactors is similar to the JET-ILW levels, wall plasma power loading should not be an issue. However, other contributions such as charge exchange may be a problem.EURATOM 63305