Configuration and engineering design of the ARIES-RS tokamak power plant

Reactor relevant ICRH scenarios have been assessed during DT experiments on the JET tokamak using H mode divertor discharges with ITER-like shapes and safety factors Deuterium minority heating in tritium plasmas was demonstrated for the first time For 9% deuterium, an ICRH power of 6 MW gave 166 MW of fusion power from reactions between suprathermal deuterons and thermal tritons The Q value of the steady state discharge reached 022 for the length of the RF flat-top (27 s), corresponding to three plasma energy replacement times The Doppler broadened neutron spectrum showed a deuteron energy of 125 keV, which was optimum for fusion and close to the critical energy Thus, strong bulk ion heating was obtained at the same time as high fusion efficiency Deuterium fractions around 20% produced the strongest ion heating together with a strong reduction of the suprathermal deuteron tail The ELMs had low amplitude and high frequency and each ELM transported less plasma energy content than the 1% required by ITER The energy confinement time, on the ITERH97-P scale, was 090, which is sufficient for ignition in ITER 3He minority heating, in approximately 50:50 D:T plasmas with up to 10% 3He, also demonstrated strong bulk ion heating Central ion temperatures up to 13 keV were achieved, together with central electron temperatures up to 12 keV The normalized H mode confinement time was 095 Second harmonic tritium heating produced energetic tritons above the critical energy This scheme heats the electrons in JET, unlike in ITER where the lower power density will allow mainly ion heating The inverted scenario of tritium minority ICRH in a deuterium plasma was demonstrated as a successful heating method producing both suprathermal neutrons and bulk ion heating Theoretical calculations of the DT reactivity mostly give excellent agreement with the measured reaction rates

https://iopscience.iop.org/article/10.1088/0029-5515/39/3/303/pdf

Bulk ion heating with ICRH in JET DT plasmas

The paper highlights the JET work in physics and technology during the period of the JET Joint Undertaking (1978-1999), with special emphasis on what has been learned for extrapolation to a NEXT STEP device. - Global confinement scaling has been extended to high currents and heating powers. Dimensionless scaling experiments of ELMy H?mode plasmas suggest that bulk plasma transport is gyro-Bohm and predict ignition for a device with ITER-FDR parameters. Experiments in which the plasma elongation and triangularity were varied independently show a strong increase of confinement time with elongation (?E ~ ?a 0.8 ? 0.3), thus supporting a basic design principle of ITER-FEAT. With the Pellet Enhanced Performance (PEP) mode, JET has discovered the beneficial effect of reversed magnetic shear on confinement, opening the possibility of advanced tokamak scenarios. - With a three stage programme of progressively more closed divertors, JET has demonstrated the benefits of divertor closure, in particular, of high divertor neutral pressure which facilitates helium removal. It has also shown that in detached (or semidetached) radiative divertor plasmas the average power load on the target plates of a NEXT STEP device should be tolerable but, in addition, that the transient power loads during ELMs could cause problems. - In 1991 JET has demonstrated the first ever controlled production of a megawatt of fusion power. More extensive D-T experiments in 1997 (DTE1) have established new records in fusion performance: 16?MW transient fusion power with Qin = 0.62 (i.e. close to breakeven, Qin = 1) and 4?MW steady state fusion power with Qin = 0.18 for 4?s. DTE1 has also allowed a successful test of various reactor ICRF heating schemes and a clear demonstration of alpha particle heating, consistent with classical expectations. - JET has developed and tested some of the most important technologies for a NEXT STEP and a reactor, in particular the safe handling of tritium and the remote handling of large equipment as demonstrated by the Remote Divertor Exchange following DTE1.

https://iopscience.iop.org/article/10.1088/0029-5515/41/12/217/pdf

The scientific success of JET

https://digital.library.unt.edu/ark:/67531/metadc664933/m2/1/high_res_d/270817.pdf

Plasma wall interaction and tritium retention in TFTR

Ion cyclotron range of frequency (ICRF) heating is foreseen as an integral component of the initial ITER operation. The status of ICRF preparations for ITER and supporting research were updated in the 2007 [Gormezano et al., Nucl. Fusion 47, S285 (2007)] report on the ITER physics basis. In this report, we summarize progress made toward the successful application of ICRF power on ITER since that time. Significant advances have been made in support of the technical design by development of new techniques for arc protection, new algorithms for tuning and matching, carrying out experimental tests of more ITER like antennas and demonstration on mockups that the design assumptions are correct. In addition, new applications of the ICRF system, beyond just bulk heating, have been proposed and explored.

/pdf/progress-on-ion-cyclotron-range-of-frequencies-heating-1bgb131klz.pdf

Progress on ion cyclotron range of frequencies heating physics and technology in support of the International Tokamak Experimental Reactor

Two new four-strap fast wave antennas, driven by 8 MW of rf source power at 30 MHz, were installed for the final run period of TFTR. These antennas were utilized with D-T target plasmas to perform experiments in mode conversion heating. Poor mode conversion efficiency seen in D-T during the 1996 run (typically ∼20% of the coupled rf power) is now believed to be due to ion cyclotron absorption by a small (∼0.5%) impurity of 7Li. This problem was remedied through the use of 6Li for wall conditioning. In contrast to the 1996 results, experiments performed with 6Li wall conditioning produced strong ion and electron heating in D-T plasmas.

Mode conversion heating and current drive in TFTR

A wide variety of potential ICRF heating scenarios relevant for the deuterium-tritium plasmas expected in tokamak reactor-class devices were explored in the TFTR and JET programs. Key physics results from the two programs are discussed.

RF physics in Deuterium-Tritium plasmas

The folded waveguide (FWG) antenna is an advanced ICRF launcher under development at ORNL that offers many significant advantages over current‐strap type antennas. These features are particularly beneficial for reactor‐relevant applications such as ITER and TPX. Previous tests of a development folded waveguide with a low density plasma load have shown a factor of 5 increase in power capability over loop antennas into similar plasma conditions. The performance and reliability of a FWG with an actual tokamak plasma load must now be verified for further acceptance of this concept. A 58 MHz, 4 MW folded waveguide is being designed and built for the PBX‐M and TFTR tokamaks at Princeton Plasma Physics Laboratory. This design has a square cross‐section that can be installed as either a fast wave (FW) or ion‐Bernstein wave (IBW) launcher by 90° rotation. Two new features of the design are: a shorter quarter‐wavelength resonator configuration and a rear‐feed input power coupling loop. Loading calculations with a standard shorting plate indicate that a launched power level of 4 MW is possible on either machine. Mechanical and disruption force analysis indicates that bolted construction will withstand the disruption loads. An experimental program is planned to characterize the plasma loading, heating effectiveness, power capability, impurity generation and other factors for both FW and IBW cases. High power tests of the new configuration are being performed with a development FWG unit on RFTF at ORNL.

/pdf/a-folded-waveguide-icrf-antenna-for-pbx-m-and-tftr-srv43ssjrf.pdf

A folded waveguide ICRF antenna for PBX-M and TFTR

NSTX is a small aspect ratio tokamak (R = 0.85 m, a = 0.65 m). The High Harmonic Fast Wave (HHFW) system is a 30 MHz, 12-element array capable of launching both symmetric and directional wave spectra for plasma heating and non-inductive current drive. It has delivered up to 6!MW for short pulses and has routinely operated at ~3 MW for 100-400 ms pulses. Results include strong, centrally-peaked electron heating in both D and He plasmas for both high and low phase velocity spectra. H-modes were obtained with application of HHFW power alone, with stored energy doubling after the L-H transition. Beta poloidal as large as unity has been obtained with significant fractions (0.4) of bootstrap current. Differences in the loop voltage are observed depending on whether the array is phased to drive current in the coor counter-current directions. A fast ion tail with energies extending up to 140 keV has been observed when HHFW interacts with 80 keV neutral beams; neutron rate and lost ion measurements, as well as modeling, indicate significant power absorption by the fast ions. Radial rf power deposition and driven current profiles have been calculated with ray tracing and kinetic full-wave codes and compared with measurements.

http://www-pub.iaea.org/MTCD/publications/PDF/csp_019c/pdf/exp2_13.pdf

RF heating and current drive on NSTX with high harmonic fast waves

Time dependent calculations of lower hybrid (LH) current profile control in Alcator C-Mod have been done using the TRANSP [1], FPPRF [2], and LSC [3] codes. Up to 3 MW of LH current drive power was applied in plasmas with high power ICRF minority heating (PICH=1.8−3 MW) and fast current ramp up. Using the experimentally measured temperature profiles, off-axis current generation resulted in nonmonotonic q-profiles with qmin≃1.6. Self-consistent effects of off-axis electron heating by the LH power were also included in the analysis and significant broadening of the electron temperature profile was found with qmin≳2 and a larger shear reversal radius.

S. Bernabei

Papers

Mode conversion heating and current drive in TFTR

RF physics in Deuterium-Tritium plasmas

A folded waveguide ICRF antenna for PBX-M and TFTR

RF heating and current drive on NSTX with high harmonic fast waves

RF current profile control studies in the alcator C-mod tokamak