C.P.C. Wong

Divertor heat and particle control experiments on the DIII-D tokamak

Abstract In this paper we present a summary of recent DIII-D divertor physics activity and plans for future divertor upgrades. During the past year, DIII-D experimental effort was focused on areas of active heat and particle control and divertor target erosion studies. Using the DIII-D Advanced Divertor system we have succeeded for the first time to control the plasma density and demonstrate helium exhaust in H-mode plasmas. Divertor heat flux control by means of D 2 gas puffing and impurity injection were studied separately and in both cases up to a factor of five reduction of the divertor peak heat flux was observed. Using the DiMES sample transfer system we have obtained erosion data on various material samples in well diagnosed plasmas and compared the results with predictions of numerical models.

Erosion and deposition of metals and carbon in the DIII-D divertor

Net erosion rates at the outer strike point of the DIII-D divertor plasma were measured for several materials during quiescent H-mode operation with deuterium plasmas. Materials examined include graphite, beryllium, tungsten, vanadium and molybdenum. For graphite, net erosion rates up to 4 nm/sec were found. Erosion rates for the metals were much smaller than for carbon. Ion fluxes from Langmuir probe measurements were used to predict gross erosion by sputtering. Measured net erosion was much smaller than predicted gross erosion. Transport of metal atoms by the plasma across the divertor surface was also examined. Light atoms were transported farther than heavy atoms as predicted by impurity transport models.

Toroidal reactor designs as a function of aspect ratio and elongation

A `common basis systems study of superconducting (SC) and normal conducting (NC) DT burning fusion power and materials testing reactor designs is presented. The figures of merit for power and materials testing reactors are, respectively, projected cost of electricity (COE) and direct cost. A common 0-D plasma modelling basis is used and the plasma geometry and engineering aspects of the SC and NC designs are treated in an equivalent manner that is consistent with the limitations of their respective magnet technologies. Aspect ratios A in the range 1.2≤A≤6 and plasma elongations κ in the range 1.5≤κ≤3 are explored and an MHD stability (beta limit) physics basis that accurately describes the increase of normalized beta βN and toroidal beta βT with decreasing A and/or increasing κ is incorporated. With this MHD basis taken into account and with the usual reactor geometry, physics and engineering constraints and costing bases applied, the results of the study show that for SC power reactor designs with κ = 2 the COE has a minimum for 2≤A≤3 and increases with a further increase in A(A>3). For NC power reactors the COE has minima at A≈2. For both SC and NC power reactors, the minima are more apparent with lower κ. While SC options appear to offer lower COE for power plants, the direct cost for NC test reactors with similar fusion power output is significantly lower. Within the NC category, test designs that combine modest A and maximum elongation show promise for achieving ITER-like testing capabilities at low direct cost. For example, an NC coil design with A = 2, κ = 3 could produce fusion power of 200 MW at 1.23 MW/m2 average neutron wall loading at a total direct cost of about 650 million US dollars. This NC design with a fissile blanket could also convert about 1270 kg of fission reactor waste per full power year. A possible cost effective development scenario for fusion power is identified for NC A = 2 toroidal devices for physics and material testing studies for use in the near future. The selection of the SC or NC coil option could then be made for the construction of the demonstration power reactor.

Helium-Cooled Blanket Designs

A systematic selection and evaluation of helium-cooled blanket concepts has been performed as part of the Blanket Comparison and Selection Study (BCSS). Helium-cooled Li/sub 2/O, lithium, LiAlO/sub 2//Be, and Flibe/Be blanket concepts were selected for detailed design and evaluation. These concepts are applicable to both tokamak and tandem mirror reactors (TMRs). The design and analysis of Li/sub 2/O, lithium, and LiAlO/sub 2//Be blanket concepts are presented. Previous blanket designs were studied and the pressurized lobe configuration was selected for the helium-cooled BCSS designs. Fifty-four different combinations of structural, breeder, and neutron multiplier materials were considered and four helium-cooled blanket concepts were selected for detailed design and evaluation. Mechanical, thermal, and neutronic designs were developed, and tritium control methods were specified. In the final BCSS evaluation, the Li/sub 2/O blanket design ranked second for tokamaks and third for TMRs. The lithium blanket design ranked third for tokamaks and fourth for TMRs. To help guide future research and development, the critical issues associated with each of the helium-cooled designs were identified and necessary experimental data highlighted. These data include irradiation behavior of the blanket materials, compatibility between the structure and liquid-metal breeder materials, and the behavior of tritium in a helium-cooled blanket morexa0» environment. The designs offer favorable performance, design simplicity, and attractive safety features for fusion reactors. Design improvements were identified that could allow still better performance of the helium-cooled blanket designs. «xa0less

ADVANCED HIGH PERFORMANCE SOLID WALL BLANKET CONCEPTS

Abstract First wall and blanket (FW/blanket) design is a crucial element in the performance and acceptance of a fusion power plant. High temperature structural and breeding materials are needed for high thermal performance. A suitable combination of structural design with the selected materials is necessary for D–T fuel sufficiency. Whenever possible, low afterheat, low chemical reactivity and low activation materials are desired to achieve passive safety and minimize the amount of high-level waste. Of course the selected fusion FW/blanket design will have to match the operational scenarios of high performance plasma. The key characteristics of eight advanced high performance FW/blanket concepts are presented in this paper. Design configurations, performance characteristics, unique advantages and issues are summarized. All reviewed designs can satisfy most of the necessary design goals. For further development, in concert with the advancement in plasma control and scrape off layer physics, additional emphasis will be needed in the areas of first wall coating material selection, design of plasma stabilization coils, consideration of reactor startup and transient events. To validate the projected performance of the advanced FW/blanket concepts the critical element is the need for 14 MeV neutron irradiation facilities for the generation of necessary engineering design data and the prediction of FW/blanket components lifetime and availability.

Characterization of energetic deuterium striking the divertor of the DIII-D tokamak

Abstract Measurements of the deuterium particle flux and energy to the divertor of the DIII-D tokamak during a series of plasmas that terminated in disruptions have been made using a silicon collector probe installed on the DiMES (divertor materials exposure system) mechanism. During the steady state portion of each discharge, the probe was located in the private flux region, but immediately before disrupting the plasma, by injecting either Ar or D2 gas, the strike point of the outer divertor leg was positioned over the probe. Comparison of the amount of retained D in the probe for the two types of disruptions indicates that much of the trapped D could have resulted from exposure in the private flux zone prior to the disruption. Measurements of the depth distribution of the trapped D in the Si imply that the incident ion energy was approximately 100 eV at normal incidence and decreased slightly at oblique angles. The measurements give an upper bound to the energy of deuterons striking the divertor floor in the vicinity of the strikepoint during disruptions.

Helium cooling of fusion reactors

Abstract On the basis of worldwide design experience and in coordination with the evolution of the International Thermonuclear Experimental Reactor (ITER) program, the application of helium as a coolant for fusion appears to be at the verge of a transition from conceptual design to engineering development. This paper presents a review of the use of helium as the coolant for fusion reactor blanket and divertor designs. The concept of a high-pressure helium cooling radial plate design was studied for both ITER and PULSAR. These designs can resolve many engineering issues, and can help with reaching the goals of low activation and high performance designs. The combination of helium cooling, advanced low-activation materials, and gas turbine technology may permit high thermal efficiency and reduced costs, resulting in the environmental advantages and competitive economics required to make fusion a 21st century power source.

Neutron wall loading of Tokamak reactors

Abstract Neutron wall loading (Γn) is a key parameter for the selection of fusion power core component materials. It also impacts the economic, performance, design, safety and environmental aspect of the fusion power plant. This paper reports the determination of the range of Γn for economically competitive fusion power plants based on the analysis that couples the MHD stability physics results to a system design code. Cost of electricity (COE) was selected as the parameter to be minimized. For both normal conducting and superconducting coil options, at thermal efficiency of 46% and at the power output range of 1–2 GW(e) the average neutron wall loading is 4–7 MW/m2. For a given power output, higher thermal efficiency will allow lower Γn. At the above range of Γn, in order to have economical fusion power reactors, for the solid first wall design option, high thermal efficiency of 46% to 57.5% requires the use of alloys like V and W-alloy, respectively. The corresponding COE can be projected to be in the economically competitive range of 62–54.6 mill/kWh.

First measurements of the ion energy distribution at the divertor strike point during DIII-D disruptions

Plasma/wall interaction studies are being carried out using the Divertor Materials Exposure System (DiMES) on DIII-D. The objective of the experiment is to determine the kinetic energy and flux of deuterium ions reaching the divertor target during argon-induced radiative disruptions. The experiment utilizes a special slotted ion analyzer mounted over a Si sample to collect the fast charge-exchange (CX) deuterium neutrals emitted within the recycled cold neutral layer (CNL) which serves as a CX target for the incident ions. A theoretical interpretation of the experiment reveals a strong forward pitch-angle dependence in the approaching ion distribution function. The depth distribution of the trapped D in the Si sample was measured using low-energy direct recoil spectroscopy. Comparison with the TRIM code using monoenergetic ions indicated that the best fit to the data was obtained for an ion energy of 100 eV. An estimate of the CNL thickness /spl int/ndl indicates that during disruptions the CNL cushion is thick enough to reduce the local ion heat load by /spl sim/30% due to CX refluxing.