ARIES Documents -- Meetings ArchiveARIES Project Meeting, 1-3 December 1999
University of Califonia, San Diego
Documented by L. Waganer
Attendees: C. Baker, E. Cheng, B. Dove, L. El-Guebaly, S. Jardin, C. Kessel, L. Lao (GA), S. Malang, T.K. Mau, R. Miller, F. Najmabadi, R. Raffray, G. Staebler (GA), D. Steiner, I. Sviatoslavsky, D-K. Sze, M. Tillack, L. Waganer, X. Wang
Encl: ARIES Action Item List
Administrative and General Topics
Bill Dove reviewed the revised FY00 Advanced Design budget, along with the anticipated program impact. Our expected $1M budget increase was reversed by Mike Holland, an OBM budget examiner. But Anne Davies designated $400 K to be available for investigation of critical IFE issues. C. Baker and F. Najmabadi briefed M. Holland on the benefits of the Advanced Design activities. [See next paragraph.] Bill mentioned that the FY01 budget request is being finalized. The budget request will include some expanded research for IFE, but the overall total may be reduced from prior years. No specific budget data are available at this time. There will be increased use of competitive initiatives involving laboratories, universities, and industry.
Farrokh Najmabadi summarized the Mike Holland briefing involving C. Baker, A. Davies, and himself. Dr. Holland was very interested in the purpose and the results of the Advanced Design studies and how they influenced the entire fusion community. The intent of the briefing was to convey that the Advanced Design studies were doing beneficial scientific work, with significant impact on the fusion sciences R&D programs. Dr. Holland was very interested and asked provocative questions.
F. Najmabadi informed the team on the restructuring of the ARIES study priorities and our immediate goals. The neutron source study must be finished (writing) by September or October 2000. The principal task leaders must finish their drafts by the end of December 1999 after the Neutron Source presentations to establish a strategy to complete the study by September 2000 (or earlier). Producing credible ARIES-AT results has become very important to the fusion community. Thus it is recommended that the timetable for the completion of the ARIES-AT design be extended to enhance the fidelity of the study results up to the expected and necessary norm. The ARIES-AT study is to be completed by June 2000 and the final report completed by the end of the FY2000. There has been much positive feedback from the community; they have been utilizing our results. Thus the ARIES team must do a good job on the AT design. The ARIES IFE effort is anticipated to commence in January 2000, probably concentrating on the target chamber.
The VLT Advisory Committee will review the ARIES study in the near future. Some support will be required for the two-day meeting.
The evolving FIRE design has many similarities with the ARIES-RS and ARIES-AT designs. Several people from the ARIES team also work on the FIRE staff. Due to the similarities and synergism of the two projects, it was resolved that ARIES and FIRE will increase their interaction through briefings and teleconferences. Also, it was recommended the next ARIES meeting in late March be held at PPPL with participation of the FIRE team. [Subsequent events may combine the US/Japan reactor workshop with the March ARIES meeting.]
The EU will commence a tokamak or stellarator power plant study in the May-June 2000 time frame, with collaboration with the ARIES team. This design group will be larger than the ARIES group and will emphasize safety and environmental effects.
The US/Japan reactor design workshop is scheduled to meet in San Diego around the September 2000 timeframe. Perhaps this workshop should be concurrent with an ARIES meeting. [See prior paragraph.]
Comparison of ARIES-AT and FIRE Designs
Chuck Kessel presented a short comparison of the physics basis and the design approach for the ARIES and the FIRE studies. Although the FIRE reactor uses passively cooled copper coils and a smaller plasma, much of the underlying physics and engineering bases are similar. Both assume a reverse shear plasma configuration with a double null divertor. The approaches for the inboard strike point are similar so it would be beneficial to collaborate on a solution. FIRE should be able to access and confirm MHD regimes necessary to progress toward ARIES-RS and ARIES-AT plasmas. The FIRE plasma will be ohmically driven to achieve 10's of seconds pulse length, whereas ARIES-AT will use on- and off-axis current drive systems to supplement bootstrap currents to achieve steady state operation. FIRE is being designed to run with and without kink stabilization shells.
An Update and Recommendations for Heating and Current Drive -Tak Kuen Mau presented his analysis results for a RF-driven current drive system for ARIES-AT using an equilibrium case for minimum current drive (CD) requirements provided by C. Kessel (8/99). The best case resulted in 8.7 MW (ICRF power) on axis and 6.2 MW (LH) off axis. He noted that the ICRF driven currents may be able to peak closer to the center of the plasma than shown. To date, his CD assessments for well-aligned equilibria have been for a zero plasma edge density. A non-zero density and a lower electron temperature at the plasma edge would imply a less efficient current drive system. His analysis results over a range of plasma edge densities were presented. It was noted that current drive efficiencies are dependent on edge temperatures, but this effect has not been included in present analyses. Results using electron-cyclotron (EC) current drive for on-axis and off-axis supplemental current drive were shown. The EC current drive system is less efficient than the LH system.
T.K. is also investigating options for plasma rotation, with neutral beam (NB) systems being the more likely option. It was suggested he assess a low energy NB option with a rigid-body plasma model to obtain a preliminary estimate of system effectiveness. The most immediate actions required are to select the system options and determine system geometries for the launchers and penetrations.
Physics Optimization of ARIES-AT - Lang Lao summarized the areas of focus on the physics optimization process as follows:
In the stability optimization, the ideal low number modes are stabilized by a conducting wall placed at r = 1.2 * A. Beta is limited by high number ideal ballooning modes near the plasma outer region. Rotational drive and radially localized off-axis current drive are essential for stabilization against resistive wall modes and neo-classical tearing modes.
In the modeling of the divertor heat load, a high radiation fraction of the total exhausted power (greater than 50%) is essential to keep peak heat flux at less than 10 MW/m2. It is necessary to accurately maintain the double-null magnetic balance within 0.5 cm.
Gary Staebler discussed the transport modeling for the ARIES-RS (presumably the analyses and results would also apply to ARIES-AT), which employs multi-code iteration. First a bootstrap-aligned MHD equilibrium was found using the code TOQ. The ONETWO transport code computed the fusion power for the equilibrium density and temperature profiles. Then the BLF23 transport model determined the steady state temperature profiles holding density and sources fixed. The fusion power was reduced to keep the pressure near the target beta. The temperature profiles were transferred to the ONETWO transport code and the fusion power was recalculated. To match up the fusion power, the DT ratio was varied along with DT fuel ratio. This process was iterated until the desired conditions were achieved. A Greenwald density limit of one was preserved in these calculations.
Farrokh was concerned that the above transport modeling procedure varied only one parameter at a time and would not yield suitable data sets in the appropriate multidimensional parameter space. Farrokh recommended that Gary start with the established ARIES-AT plasma profiles and scaling from the systems code to establish the parameter space. The Greenwald density limit will probably be exceeded, e.g. the current ARIES-AT (1 GW) strawman has a density of 1.62 times the Greenwald limit.
Lang Lao discussed the non-inductive current ramp up using a 0-D analysis with fixed profiles and geometry. A physics-based transport model was used for the internal transport barrier (ITB) analysis. Current drive and rotational flow drive used fast wave, electron cyclotron waves and negative ion beams. Improved heat and particle controls were included. These effects will be included in an integrated model with self-consistent current and transport codes to study bootstrap alignment, ITB sustainment, and stable paths to high beta and bootstrap fraction operation. A rampup of approximately 1 hour would be required to maintain a dI/dt greater than zero. The base case requires a neutral beam power of 150 MW. This power level can be reduced to approximately 30 MW with better confinement and control (H98y1 ~2).
Extension of the RS Physics Basis to ARIES-AT - Chuck Kessel described the analysis improvements over the previous ARIES-RS basis. The 99% flux surface is used rather than the 95% flux surface used on ARIES-RS. The 99% flux surface would correspond to larger plasma elongation and triangularity, which yields a higher stable beta. A more flexible pressure profile formulation allows better bootstrap alignment and higher ballooning beta limit. With the higher triangularity, the inboard slot is virtually eliminated, which allows higher betaN, higher plasma currents, and higher beta. However, the lack of or minimization of the inboard slot results in a severe impact to the design of the inboard divertor region. Placing the vertical stabilizing shell closer to the plasma allows for higher elongation, thus increasing the betaN, plasma current, and beta. The increased plasma elongation allows the vertical stabilizing shell to be placed closer to the plasma, thus increasing the betaN, plasma current, and beta. The HHFW can be eliminated and alternative off-axis current drive systems can be used. At present, the most likely stable plasma configuration will have an elongation of 2.2 and a beta of 11.6%.
Chuck addressed the effect of a finite edge plasma density. An increased edge density will allow additional radiation from the edge, hence a cooler edge temperature. In the ARIES-RS analysis, an edge density equivalent to 40% of the central plasma density was considered, which would relieve the divertor heat load. However, this strongly reduced the bootstrap current, increased the CD requirement at the plasma edge, and increased the plasma core Zeff, further increasing the CD power required. As a compromise, the edge density was reduced to 20% with an added neon impurity being employed as the radiating impurity with a Zeff = 2.0. For ARIES-AT betaN cases of 5.6, 6.0, and 6.5 with stable equilibria, the off-axis CD requirement has increased to 1.2 MA.
Most tokamak experimental plasmas are limited by the Greenwald density limit. However, high beta tokamak power plants will probably require a plasma density higher than the Greenwald density limit. Recent experiments, such as TEXTOR and DIII-D, have indicated that this limitation is related to unstable MARFE formation, but it may be avoided up to twice the Greenwald density limit. This would suggest that the ARIES-AT plasma must operate with an edge temperature greater than 100 eV and the inboard strike points must have efficient pumping of neutrals. These conditions are to be tested and validated on DIII-D in the future.
Chuck discussed his approach for constraining the plasma density and temperature profiles, mainly to optimize bootstrap current and MHD stability. The assumed profiles are similar to experimental results, but the gradients are spread out more than the observed profiles. This could be accomplished with some form of internal transport barrier (ITB) control to relax and spread the gradients. Density and temperature profiles outside the qmin are important, affecting the divertor design, current drive requirements, and neoclassical tearing stability.
Neutron Source Study
Status of Neutron Source Study - Don Steiner emphasized that we should be referring to candidate neutron source applications as "intermediate goals" rather than "near-term" as these applications will not be realized in the near term. He addressed the advantages of a catalyzed DD system as requiring no tritium breeding, reduced divertor heat fluxes, and reduced neutron damage.
Don discussed the need to identify the products and customers, elucidate the clear advantages for fusion, and to quantify the metrics for fusion and competitors. Les Waganer has completed the market assessment in draft form. Dave Petti has completed the preliminary safety and environment review.
System Studies for the Neutron Source Study -Ron Miller related that he had completed a preliminary assessment based upon the ARIES-ST tokamak physics (ARIES-NS). There was a lot of discussion that a high beta case may not be appropriate or correct in this use. Also the cost of electricity may not be the best measure of merit as the capital and operating cost of a processing facility has not been included at this time.
Update on Pu and Actinide Transmutation in Fusion Reactors - Ed Cheng described the features of the HTBR-type transmutation blanket. The solid blanket contained the Pu/actinides in graphite/SiC coated particles within a helium-cooled graphite matrix. There is an option that allows shuffling of the fuel elements to different zones to maximize the burnup-fraction to approximately 90%. He explained that the ST-VNS transmutation reactor had four regions to enhance burnup.
A molten salt blanket is being evaluated for an IFE reactor. It uses a self-cooled molten salt with 0.5 molar % Pu/actinide fluorine. Four distinct zones are used to enhance burnup. Japan is starting an assessment of a molten salt transmutation reactor.
Engineering and Nuclear Performance Assessment -Mark Tillack summarized his work on compiling and assessing the engineering and nuclear performance of various neutron source applications, including fission, fusion, and accelerators. He is focusing his efforts on fission waste and weapons Pu destruction. Within that scope, he is further focussing on performance, not feasibility, safety, or operations. The blanket and the fuel cycle dominate most of the performance parameters. He reviewed the feedstock, disposition scenario, and processing mode options. Within each of the option types (fission, fusion, and accelerators), Mark summarized the complete matrix of options including type, coolant/moderator, fuel, and breeder/target.
Mark reviewed the list of parameters being considered, highlighting two examples:
Key Neutronic Performance Parameters
Engineering Performance Parameters
Mark concluded that:
Neutron Studies Wrap-up - Don Steiner summarized the major differences in the blanket approaches that lead to diverse solutions and even different metrics. At present there are no common metrics to easily compare different products and approaches. If burnup of Pu were an important metric, then accelerator-based systems would be favored. Don suggested more systems studies to investigate catalyzed DD systems with low engineering Q values to quantitatively compare with the DT tokamak systems.
Farrokh Najmabadi wanted the ability to better assess the neutron study current status - he suggested all uncompleted and undocumented study elements be written up in draft form and forwarded to Don and himself by the end of December 1999. This would enable an assessment of progress and the likelihood of a reasonable conclusion of Phase I (Concept Definition Phase). Les Waganer suggested a flow chart outlining the requirements, concepts, advantages, disadvantages, decision points, and development pathways. Charlie Baker suggested a list of advantages (and disadvantages) for each approach along with the concept rationale.
Ron Miller highlighted the main features of ARIES-AT: higher performance physics (~50% higher beta), higher performance magnets (not fully utilized at present), higher performance blanket (thermal conversion efficiency ~60% vs. 46%), high performance compact shield, and reduced component costs (HTS magnet structures). Positive improvements in plasma physics and engineering had yielded steady improvements in the viability of electrical generating fusion power plants since the 1980's.
Ron detailed the ARIES-AT data inputs and inherent assumptions contained in the current ARIES system code. The basis for the HTS magnet cost improvements must be justified in detail. The current drive systems must be determined and appropriate scaling and power requirements must be specified. The additional cost to achieve the higher thermal efficiency must be quantified. The cost basis is being updated from 1992 dollars to 1999 dollars, along with other accounting changes in line with those proposed by Delene.
Farrokh recommended that Ron conduct parametric trade studies to determine the optimum 1.0 GW power plant that will become our next strawman. Higher power plants will only be defined from the ARIES code, with no detailed design effort on the larger plants.
Mark Tillack presented Dave Petti's Loss of Vacuum Analysis (LOVA). The risk dominant events are based on ITER assessments associated with bypass of radiological confinement barriers. He assumed a failure of double confinement in a heating and current drive system line, causing air to enter the plasma chamber and a plasma disruption, which mobilizes the dust and tritium. The air exchanges between the chamber and the "generic bypass" room just outside the chamber. Several cases were studied with the MELCORE integrated computer code to determine sensitivities to LOVA assumptions. Tungsten and tritium releases for a 0.02 m2 break are rather insensitive to geometry and leakage and isolation dynamics. The heavy tungsten dust settled rather quickly with minimal dispersion. Tungsten dust was less than 67% of the allowed limit. The tritium releases should be tolerable so long as the mobile inventory is less than 3 kg HTO. (~450 g of tritium).
ARIES-AT Nuclear Parameters, Radial Build, and Activation Analyses - Laila El-Guebaly summarized the ARIES-AT power core design parameters, noting changes from ARIES-RS and previous assumptions regarding the AT design basis. With the presently assumed configuration and materials selection (SiC, 90% enriched LiPb, partial tungsten vertical stabilizing shells, and necessary gaps and penetrations), the nuclear parameter requirements are just achieved. Increases in SiC, stabilizing shells, and penetrations/gaps will reduce the tritium breeding, while a thicker blanket will increase the breeding. Enrichment of the lithium above 90% would not be cost effective. Efforts to validate the enrichment costs are being pursued. The tungsten stabilizing shells significantly reduce the breeding behind these elements (10% of breeding is behind these shells and the 6-cm-thick W shell reduces that breeding by 50%, that is a 5% reduction overall), so material and design changes are being evaluated.
The lifetime of power core components is based on 3% burnup of SiC and a 200 dpa for ferritic steel in the shielding and vacuum vessel. All power core components except for the first wall, divertor surfaces, and the innermost blanket modules are life of plant components (40 FPY). The nuclear heat load to all in-vessel was presented based upon the overall neutron energy multiplication of 1.1.
Laila presented the inboard and outboard radial builds to achieve the necessary nuclear shielding requirement, heat generation, and tritium breeding. Tradeoff analyses of the water and filler materials of the LT shield and vacuum vessel (inboard and outboard) was conducted. Farrokh suggested operating at an off-optimum design point and removal of the WC and FS fillers in the vacuum vessel. L. Waganer suggested that design considerations would likely require higher water content (>10%) for the inboard shield. Laila will assess the impact of those changes on the radial build. A comparison between ARIES-AT and ARIES-RS showed 17 and 35 cm reduction in the IB and OB radial builds, respectively, due to the better shielding performance of the LiPb breeder and the ability to use water in the shield and vacuum vessel.
Activation analysis results were given for the outboard power core elements. The inboard component results would be similar, but at reduced levels. The outboard backwall of the inner blanket has much lower activity and decay heat than the first wall. Conceivably, the outboard wall could be recycled and reused if deemed cost effective. The outboard first wall and blanket could easily qualify as class C low-level waste after its life of 3 FPY. Similar trends were seen in the outboard, outer blanket, high temperature shield, vacuum vessel, and magnet, which allowed these components to qualify as class C low-level waste after a life of 40 FPY.
Preliminary Design Definition and Integration of Power Core - Les Waganer noted that it was necessary to define the function and geometry of the power core components to assess the feasibility of applying low cost techniques to the vacuum vessel and low temperature shield. Based upon the general plasma size and shape along with the initial radial build assumptions, Les constructed a schematic representation of the inboard and outboard power core elements including modifying TF coils with the outer legs further outboard. The maintenance approach was assumed to be similar to ARIES-RS. (The next step is to develop a detailed CAD model of the power core to verify component sizes and maintenance approach.) With 16 sectors, it appeared that the life-limited component modules could be withdrawn between the outer TF coil legs. He also suggested that the inboard low temperature shield and vacuum vessel be combined into a single unit due to similar lifetimes, common materials, and structural connections. This would also simplify plumbing connections on the inboard region and increase the structural rigidity of the vacuum vessel. (Laila made the same suggestion at the May 1999 meeting - see her presentation of that date. The problem is that if there is a need to cut/weld the VV/shield, the welds must be protected. At present, the inner surface of the LT shield does not meet the reweldability limit at all poloidal locations).
Les examined approaches to achieve the recommended composition of ferritic steel, water, and WC in the inboard low temperature shield. He concluded that an overall water fraction of 5% and a WC fraction of 80% (local fraction of 6% and 94%, respectively) would not be a feasible solution. A higher water fraction was recommended by Les. Laila El-Guebaly agreed to examine alternative solutions along these lines. Les outlined an initial design approach for fabrication and assembly of the vacuum vessel and low temperature shield.
Initial Considerations on Core Configuration and Design Integration - Mark Tillack discussed the complexity of the vacuum vessel approach suggested by Waganer (ala ARIES-RS). He would like to examine an approach with removal of smaller components through smaller and fewer port openings. However, he recognizes that this would imply a more complex in-vessel manipulator and handling system. This approach might have an adverse effect on the scheduled and unscheduled maintenance times. He will be working with Xueren Wang to define a CAD model of the power core to examine these issues. This model will provide more definitive information on the vacuum vessel and other systems for higher fidelity engineering models.
Status of ARIES-AT Blanket and First Wall Design - Rene Raffray noted that ORNL is sponsoring a SiC town meeting in mid January 2000. He and several other ARIES team members will be participating in this meeting.
Rene presented the evolving first wall and blanket design, along with a table of SiC/SiC properties. He also presented the current AT machine and power parameters used in his analysis. The average/maximum heat flux is 0.51/0.71 MW/m2 and the average/maximum neutron wall loads are 5.2/6.1 and 3.0/4.0 for outboard and inboard regions, respectively. He noted that one of the principal reasons for selecting SiC is the ability to achieve high thermal conversion efficiency as shown in the thermal conversion cycle schematic and cycle analyses.
He showed a three-layer blanket with all internal rib structures. The blanket structure is first cooled with the LiPb coolant and then the coolant is passed through the larger internal channels where the final, high exit temperature is achieved. The cooling of the structure first allows it to remain below the 1000°C structural temperature limit, yet the bulk coolant can achieve the desired 1100°C outlet temperature. Iteration of the channel sizes and configurations in the first wall and structure were conducted to maximize heat transfer into the coolant with a minimum pressure drop. He was particularly concerned about the channel flow characteristics to predict the heat transfer from the wall into the fluid, including velocity profile effects and MHD flow laminarization. He is concerned about the region where the coolant flow makes a right angle turn and become parallel to the magnetic field; flow might stagnate at that area. This could be alleviated with an inertially dominated, high velocity flow in that region.
Update on First Wall and Blanket Design - Igor Sviatoslavsky has been concentrating on the coolant routing in the first wall and blanket elements. Tradeoffs of the tube spacing or conversion from round to elliptical tubes were discussed. The elliptical shape has much higher stresses. A complete routing of the coolant was shown for all elements. Toroidal routing is much more complex. Thermal stress calculations were accomplished on the poloidal and toroidal flow tubes, with the toroidal tube having the higher stresses.
A static load stress calculation was conducted to determine the stresses on the freestanding blanket module. Waganer suggested that rather than a half-height model, the entire model should be modeled since the loading of the part is not symmetrical about the midplane and the maximum stresses will probably occur at the bottom.
Alternative Geometry for Self-Cooled Blanket - Siegfried Malang presented a new design approach for the blanket that builds upon the attractive features of the baseline blanket (SiC structure, LiPb coolant, higher power density, high temperature operation, and radial segmentation). His goal is to reduce the MHD effects and lower the fabrication costs.
The proposed design is a long curved structure with coolant flow from top to bottom and back to the top. The design employs concentric flow passages having the cooler flow through the outer regions and a slower flow in the central region to obtain the final high temperature exit fluid. The inner tube is internally smooth with many external ribs perpendicular to the surface. Over the ribbed inner structure, a split, smooth jacket will closely fit over the inner tube to create a series of channels around the surface of the inner tube. The two elements are not structurally attached. The channels can either continuously spiral around the inner tube to distribute the heat evenly in all channels or the channels can run straight and are sized for higher velocity in the higher heating zones. A preliminary strength analysis confirms the design is within design allowables. The MHD effects are much improved. A preliminary fabrication sequence was presented for the long elements, end plenums, and plumbing.
Divertor Options and Considerations - Rene Raffray is just commencing on the divertor design. The major issues are the anticipated surface heat erosion and the high surface heat flux. He would like to retain SiC as the structural material and LiPb as the coolant, based on commonality with the remainder of the power core components. Incorporation of tungsten is being considered. Dry wall concepts are being developed. The ALPS project (R. Mattas) is being solicited to help provide a liquid metal divertor design that will meet the ARIES-AT design requirements. Farrokh Najmabadi is requested to help determine the plasma edge conditions and interaction with the divertor surfaces. Also the amount of power delivered to the inboard and outboard divertor regions need to be defined. This is especially critical for the inboard divertor area, as there is insufficient space in this region to provide a slot divertor. Instead, there probably will just be a strikeplate to absorb the surface heat and particle flux with sufficient pumping of neutrals in the near vicinity.
Rene showed a preliminary divertor element with a high velocity region to intercept the divertor surface heat flux. Especially critical is the flow region where the coolant must accomplish a right angle turn to be aligned with the magnetic field. Again, he is hoping inertial flow will help provide adequate flow velocity and heat transmission in that area.
Assessment of Tritium Handling and Storage Systems - Dai-Kai Sze outlined the primary blanket tritium issues: inventory and partial pressure limits. Recovery methods are available to maintain tritium within the inventory limit, but the partial pressure limit is more difficult to achieve with some of the anticipated blanket materials. Recovery from LiPb, which has low tritium solubility, must be a very efficient process to achieve the desired partial pressure limit. It is expected that tritium will have a low solubility and diffusivity in SiC, hence the tritium partial pressure will likely be higher than a conventional fusion blanket.
Dai-Kai presented a set of blanket tritium system parameter data for the proposed design approach. The value of the preliminary tritium partial pressure of 0.16 Pa/pass is too high to be acceptable for conventional materials and it may be too high even with SiC. The allowable tritium partial pressure is assumed to be 16 Pa. A tritium recovery efficiency of 10% was assumed.
Based upon CFFTP analyses for ITER, the coolant inventory would be approximately 0.08 m3/MW or 200 m3 of LiPb for ARIES-AT. With a tritium concentration of 1.35 wppb, the total tritium inventory in the LiPb would be 0.04 g.
Dai-Kai concluded that the low tritium solubility in LiPb requires a very efficient recovery system to process the entire coolant stream with each pass. Even with these assumptions, the tritium partial pressure will be quite high (~ 10 Pa). The experimental data for beta SiC would suggest a low tritium solubility and diffusivity. Radiation damage may increase both of these parameters.
ARIES 1-3 December Meeting Action Items
Configuration and Maintenance Approach
Neutronics and Shielding
Heating and Current Drive
PF and TF Magnets
Fuel Cycle System and Main Heat Transfer and Transport System
Vertical Stabilizing Shell
Safety and Waste Management