Guests: V. Chan, G. Voss (Culham)Attachments: ARIES Meeting Agenda (4-6 May 1998)
Farrokh Najmabadi discussed an upcoming distribution of a draft plan for the Fusion Technology Virtual Laboratory. There are also several white papers being prepared including Advanced Design areas to be assessed. The ARIES-ST project is progressing toward completion this year. Plans for next year's investigations are being considered, with a project discussion planned for this meeting (see last section of meeting minutes).
The next ARIES meeting is to be held at ANL on of 6 and 7 August. Dai-Kai Sze is in charge of the meeting arrangements.
Bill Dove informed the group that DOE management and the fusion community in general have a high regard for the assessment and design work conducted by the ARIES team. Areas of positive mention included physics and design thoroughness, assessment of innovative concepts, and safety assessments.
Bill Dove spoke in regard to the upcoming work areas for the ARIES team. He felt 1998 would be a transition period, finishing up ARIES-ST project and preparing for new topics of investigation. The recommended funding for FY'99 has not been sent to Congress. Bill Dove thought there might be multiple topics being addressed by various elements of the ARIES team. There is budget pressure to reduce the general fusion budget, but there remains general support of fusion energy development. The FY2000 budget is under pressure to reduce. Possible new topic areas for the ARIES team include an intense neutron source study (for multi-purpose use, including commercial applications), an in-depth stellarator design study (perhaps a proof-of-principle (POP)), or perhaps other areas. Small studies are also being proposed by other groups, such as a POP magnetized target designs, RFP reactor plant design, and a couple of field reversed confinement concepts.
Laila El-Guebaly summarized the US planning meeting for Major Next Step Fusion Experiments that was held last week in Madison, Wisconsin. The overall goal was to define an affordable fusion experimental facility (or facilities) that would fulfill a meaningful and beneficial mission. Possible concepts or approaches included: (a) a reduced cost ITER approach with cost-reducing innovations and lessened mission goals, (b) a combination of much smaller, single-purpose experimental machines, and (c) deferring mission goal achievement until a more attractive fusion confinement concept is verified. Candidates for the latter include the FRC, stellarator, advanced tokamak, spherical tokamak, RFP, or IFE. The decision on a recommended approach is scheduled within the next few months.
Ron Miller noted that he continues to upgrade the ARIES System Code (ASC) with new physics, engineering, and economic data and modeling algorithms. Pre-strawman solution data continue to be developed and distributed to the ARIES team to help assess key assumptions and parametric data. Data set J was chosen at the 13 April conference call as the common data set to be used at this meeting for comparative purposes as a common baseline. It is recognized as a case with a low beta and high recirculating power fraction. The neutron wall loading (NWL) is not a const raining variable but the surface heat flux may be constraining. Some attractive plasma and design changes have been identified but not yet incorporated, e.g. the flared centerpost (CP), a thinner inboard shield, placement of the divertor PF coils inside the TF coils, and reduction of the inboard scrapeoff thickness. The copper filling fraction of the CP will be increased to 85% to help reduce the CP joule losses.
Ron noted that the relatively low beta of this Case J contributes to the poor performance of this case. Throughout the meeting, plasma parameter changes were identified and discussed to improve the plasma performance. As soon as these improvements are verified, Ron will incorporate them into a new strawman case.
A limited set of aspect ratios (A = 1.25, 1.4, 1.6, and 1.8) have been selected where the Physics Group conducted numerous detailed equilibrium and stability analysis. The results of these data sets suggested the more attractive economic aspect was likely A = 1.6. But the coarseness of the aspect ratio data sets has prevented determination of the true optimum point at some intermediate aspect ratio. During the physics discussions, it was recommended the A = 1.4 case be reassessed with the more current improvements associated with the A = 1.6 case being applied to the A = 1.4 case. This might result in a more attractive case, with a lower plasma current and lower TF coil currents. If the results from this A = 1.4 investigation would suggest an intermediate point might be more appropriate, then a limited investigation of an intermediate case might be recommended.
Noting the low gradient of COE relative to increases in minor radius, it was suggested that the machine size be increased to relieve the engineering constraints with a minimal impact on the COE.
Ron is planning to update the costing basis from 1992 dollars to 1996 dollars. He discussed the economic basis and groundrules that affect the fixed charge rate (FCR). Farrokh Najmabadi questioned Ron's statement that the FCR has increased while the interest rate during the same time period has lowered [R. Miller - Need to better understand the influencing parameters].
Design Integration, Configuration, and Maintenance- Mark Tillack noted that the configuration presented during this meeting is based on the ASC Case J. It is his intent to have final drawings of the baseline in time for the next meeting.
The upper and lower tungsten stabilizing shells have been added between the LiPb portion of the blanket and the helium manifold region in the outboard blanket. The design as shown has both inboard and outboard divertor slots and transport power handling surfaces. The physics group now believes so little power would be intercepted in the inboard slot that it should be closed up and not function as a divertor. The plasma would remain as a diverted plasma but the inboard transport power (~10% of the total) would be spread over the inboard shield region [distribution TBD by GA] . A nominal inboard scrapeoff layer (SOL) thickness would be retained; but the inboard low field gradient and the lack of an inboard PF coil renders this SOL region to be indeterminate. Also the intimate presence of the outer flux surface of the plasma indicates a need for an armor material for some or all of the inboard shield surface.
Laila El-Guebaly had previously suggested thinning the upper and lower inboard shield thickness to 10 cm, which would provide a limited volume for CP flaring. Elimination of the inboard divertor slot will provide additional volume for CP flaring. CP flaring is needed to help reduce the CP dissipation power (presently around 710 MW for Case J).
The present outboard divertor slot is 2 meters long; but to help decrease the power core height, including the CP length, Mark Tillack agreed to reduce the length of the slot to 1.8 m if the vertical height of the upper shielding thickness can be reduced 20 cm by using the more efficient WC shielding material. The vacuum duct will emanate from the outer slot and extend outward rather than upward on the upper region. Tungsten will likely be added in the divertors as a plasma facing material. Impurity gas will be added in the divertor slot to enhance local radiation and spread the transport power.
Mark illustrated the major power core elements and the assembly and maintenance sequences to remove and install these elements. He noted an improved power core support mechanism proposed by Les Waganer. The group questioned the removal of the low temperature shield during routine maintenance since a previous decision had been reached to keep the low temperature shield as a permanent structure. Keeping the low temperature shield as a permanent structure would also reduce the number of coolant piping disconnects.
Details on the routing of all major plumbing were shown to determine the most efficient paths and accommodate maintenance. Two efficient helium cooling approaches to improve the heat transfer to the cooling medium for the high surface flux regions were described, namely a slotted duct and porous metal. L. Waganer showed test specimens of tungsten rods embedded in a water cooled copper substrate that were successfully tested to 5 MW/m2 for a series of 1000 cycles.
Because Boeing (Driemeyer, et al.) has collected an extensive database for fabricating large copper structures, it was suggested they might help improve the CP cost estimate. L. Waganer and W. Reiersen will coordinate this activity.
PF and TF Magnet Design Studies - Wayne Reiersen reflected that the divertor space dictates the ability to flare the CP. The incentive to flare the CP is significant in regard to the recirculating power and overall system economies. Any opportunities for improvement in this area will be explored. Wayne will investigate an idea to ease maintenance of the CP shield with a flared CP.
The return TF legs are formed as a continuous shell with a thickness on midplane of approximately 0.5 m for Al conductor. The shape is determined to keep a constant stress in the outer shell whereas the thickness is determined to keep a constant cross-sectional area. The shell will also function as the vacuum vessel. The low temperature shield will be attached to the interior of the TF coil assembly to meet the radiation limits for the shell. The conductor material can either be copper, like the centerpost, or aluminum. If it is aluminum, no low temperature shield will be required for this component.
The busbars will be moved to the midplane of the power core with an electrical insulating break on the centerline. The demountable joint will be below the midplane as required for maintenance. The outside shield will be separately supported during maintenance; but during operation, the entire TF coil set will be supported as a unit from below. A sliding joint is planned at the bottom of the CP. Alternative joint concepts being considered are FELTMETAL®, liquid metal, and sliding contacts.
All the PF coils are superconducting (SC). The two upper and two lower divertor PF coils are to be located inside the TF coils, whereas the remainder of the PF coils are located outside the TF coil set. Sufficient space will be provided for cryogenic cases around the SC coils - 5 cm was suggested as being adequate. A supporting structure should also be added and shown in the drawings.
Using the base Case J, the joule losses in the straight CP is 710 MW with cold water cooling (recommended approach) and 825 MW for warm water cooling. The nuclear heating in the CP is 200 MW. All coil and CP cooling parameters are reasonable. The mechanical stresses are < 50% of the tensile yield stress allowables for ductile copper, but the brittle copper allowables are TBD.Results of the Engineering Workshop Comparing the ARIES-ST and Culham Reactor Magnet Systems - Wayne Reisersen noted the many similarities of the Culham and ARIES-ST TF and PF magnet systems. He highlighted a few of the differences that arose from the two designs, depending on some basic plasma parameters and engineering choices. Both had a CP shield, but the choice to eliminate beryllium from the ARIES blanket resulted in a lower tritium breeding ratio (TBR). The ARIES team selected a blanket configuration and calculated the TBR to be marginal. Culham has not selected their blanket configuration and they believe they have significant margin on the TBR, particularly with a solid breeder option and a beryllium multiplier. This lower TBR determined a thicker CP shield design (20 cm) without water cooling. The Culham design with a higher TBR can use a thinner, water-cooled shield (10 cm). M. Billone cautioned that it would not be good practice to mix Be and Li2O pebbles together in the same blanket bed as is being planned for the Culham blanket design. The Culham design is not required to meet the requirement of Class C waste disposal, so a higher level of induced radioactivity can be tolerated at the end of the component lifetime; hence, a thinner shield can be used.
The level of joule losses in the CP (only) were significantly higher in ARIES-ST than the Culham design (710 MW vs 165 MW). The Culham centerpost is a flared design, but the present design of ARIES-ST is a straight centerpost, which certainly contributes to the higher joule losses. Although the aspect ratio is lower for the Culham design (1.4 vs 1.6), the machines are very similar in size and shape. However the ratio of the TF current to the plasma current for the ARIES-ST design is significantly higher at 1.43 versus 1.05. Since the joule losses are proportional to the square of the current, the ratio of the joule losses is 1.85:1. Another difference is that ARIES chose a copper CP while Culham is considering both a low-temperature aluminum and a copper CP. Previously, the ARIES team had assessed the low temperature aluminum option; but when the cryoplant efficiency and cost were considered, it was less attractive than the normal temperature copper CP. Culham is quoting a much more efficient cryoplant, in the range of 18W/W overall efficiency. The cost of the Culham cryoplant has not been evaluated yet. The Culham design uses 16 discrete TF return legs, while the ARIES design uses a uniform shell return leg arrangement. While this arrangement appears quite different, electrically there is little difference.
As a result of this comparison, the ARIES team will re-evaluate the efficiency of the cryoplant and revisit the lower aspect ratio plasma to improve the joule losses in the CP. Culham will investigate the appropriateness of mixing Be and and Li2O pebbles together in the same blanket bed.
PF Coil Design- The baseline design has assumed all PF coils are superconducting; but in response to an earlier question, Leslie Bromberg reported that, if the two upper and two lower divertor coils were normally conducting copper coils, they would have approximately 500 MW of joule heating, which would be unacceptable. So all PF coils remain as superconducting. Due to their field strength, the PF1 and PF2 divertor coils (two upper and two lower) should be Nb3Sn and the remaining SC coils, outside the TF coils can be NbTi coils. They also may be combined into as single coil if it proves to be advantageous. Leslie determined all the coil parameters for the Case J PF coils including a complete costing analysis. The total cost of the first-of-a-kind PF coil set would be $208M using ITER costing rules. Ron Miller will use this data to update the ASC PF modeling algorithms.
High Critical Temperature Superconductor Usage- Due to Leslie Bromberg's ongoing assessment of high critical temperature superconductors, he recommended consideration of this class of materials for the centerpost conductor. There are two high temperature superconducting (HTS) materials, YBCO (YBa2Cu3O7-x) and BSCCO (Bi2Sr2CaCu2O8+y ) (B-2212), to be considered. The YBCO form that Leslie was describing is a highly textured tape. There is no quench protection needed as there is no stabilizer. But if it quenches, the conductor will self-destroy itself rapidly. To achieve the high levels of critical current density, this tape should be aligned such that the surface of the tape is parallel to B (c perpendicular to B). This can be accomplished with a simple twisting of the tape. MIT has been using BSCCO 2212 in magnets with record fields of 3 T @ 4K and 2 T @ 20K. Boeing is using a wire form of the YBCO material they have developed in a DOE-funded 10 kW flywheel energy storage system with HTS bearings. The wire form may not have the geometry constraint of the tape.
These high temperature superconducting materials rely on some conductivity enhancement produced by particle beam irradiation. But excessive irradiation will degrade performance; thus the HTS materials may require more shielding than the Nb3Sn and the NbTi materials. Their advantages are that they can produce much higher fields with lower power requirements for cooling, probably LN2 or He. Leslie will continue to evaluate the application of these materials for the spherical tokamak centerpost.
Blanket Design and Analysis Summary- To better handle the outboard surface heat flux, Dai-Kai Sze proposed modifying some of the first wall coolant parameters as shown in the table below. The assumed distribution of transport heat is 50% to the first walls and 50% to the divertors. The MHD pumping power for the LiPb loop is to be calculated.
|Coolant Inlet Temperature||350oC||300oC|
|Coolant Outlet Temperature||450oC||500oC|
|Maximum Structural Temperature||550oC||600oC|
|First Wall Thickness||4 mm||3 mm|
Cooling the inboard shield is much more difficult because the radial space is very limited and the shield has a long flow path, a high surface heat flux, and a large nuclear heating. The following coolants were rejected because: liquid metals has severe MHD effects, H2O absorbs the neutrons (reducing the TBR), and organics will decompose. Helium is possible but it must be at a high pressure and large flow rates, which suggests high pumping power. Dai-Kai showed a novel approach to reduce the flow length of the high velocity region, but it does require a few (valuable) centimeters for the plenum. Don Steiner suggested the use of D2O because deuterium will not absorb the neutrons like hydrogen does. Dai-Kai and Laila will investigate this option. The cooling capability with D2O should be improved over helium with reduced pumping power and flow velocity.
Dai-Kai reported that more recent data from R. Causey indicates three orders of magnitude decrease in tritium solubility in SiC in the temperature range of interest. With this level of solubility, retention of tritium in SiC should not be a problem in the blanket. F. Najmabadi requested an estimate of the tritium inventory in the plant, by component. Dai-Kai will also define a tritium recovery and cleanup system
One of the other problems in the heat transfer and transport systems is choice of the heat exchanger material to be compatible with the heat transfer fluids, their operating temperatures, and the containment of tritium. Dai-Kai proposed SiC, W, and Mo alloys, with the SiC being suggested as the best option. L. Waganer suggested building the HX with a more conventional metal that has good strength and conductivity characteristics at the expected operating conditions. Then, this structural and heat transfer material would be coated with a corrosion-resistant and tritium-permeation barrier material. Mike Billone will offer suggestions on material choices for the HX.
The technical basis for the baseline thermal conversion cycle needs to be updated to incorporate the latest innovations and still have a credible design. L. Bromberg volunteered to provide D-K Sze with the latest information on cost and efficiency of recuperators. L. Bromberg offered performance and cost data on the latest closed-cycle helium thermal conversion systems. The ASC cost basis is not consistent with these types of thermal conversion systems and needs to be updated. Dai-Kai is also planning to define the use of the low-grade heat to make the overall plant more efficient and economical.
Breeding and Shielding Issues - Laila discussed the differences between EU-MANET, the EU-EUROFER, and the US ORNL 9Cr-2WVTa ferritic steels. Approximately 1% tungsten is needed for strength in the ferritic steels (in lieu of other less desirable constituents). Higher W (2%) reduces the TBR of solid breeder blankets by a few percentage, but does not change the breeding level of LiPb blankets. The ferritic steel for ARIES-ST is the ORNL 9Cr-2WVTa steel with 2% tungsten, which yields an acceptable TBR in our LiPb blanket design. The present design basis has an acceptable tritium breeding ratio of approximately 1.1, which will be confirmed by a 3-D analysis in the future. The RF penetrations only lower the TBR by ~2%, and the use of a neutral beam system will be even less. Laila also illustrated the power split between the He and the LiPb heat transfer loops as 48% and 52%, respectively.
Laila noted that the inboard shield can be reduced in thickness to ~10 cm above (and below) the X-point due to low NWL in this region. She showed the radial (or vertical) builds for the inboard, outboard or divertor regions. Outboard, the tungsten vertical stability shell will be added to the material build. A trade study may be needed to compare the copper and aluminum outer TF coil shell. Inboard, the two distinct regions may be required to discriminate the thinner shields in the divertor region along with the new "one-limbed" divertor region. Also the effect of the D2O inboard shield coolant will be analyzed for possible incorporation. In the divertor region extending upward (or downward), the build will be modified to include the PF coils inside the TF coils, addition of WC shield material, and rearrangement of the vacuum ducts.
Safety Analysis Results - Hesham Khater reviewed the basic machine assumptions as related to the safety analysis. The average NWL (based on a 3-D analysis) is 1.83 MW/m2 for the inboard region and is 5.14 MW/m2 for the outboard region. The ferritic steel composition was assumed to be the ORNL ferritic steel of 9Cr-2WVTa with a reduced 0.5 wppm niobium impurity level. Waste disposal ratings (WDR) were calculated, both using the NRC 10CFR61 and Fetter limits for compacted wastes one year after shutdown or removal. The removal of the niobium impurity in the ferritic steel enabled achievement of Class C waste rating (WDR < 1.0). The highest WDR of 0.83 for the CP was derived from the transmutation of Cu into 63Ni. Class C is also achieved using the Fetter limits with the main contributors being 108Ag, 192Ir, and 94Nb.
Even with a shielded CP, the copper centerpost will transmute into Co, Ni, and Zn in the reactor environment. This effect is most pronounced at the outer radii of the CP. The change in copper resistivity is dominated by the production of the nickel isotopes 64Ni and 62Ni. The transmutation-induced resistivity scales somewhat linearly with neutron fluence.
The major sources of radiological hazards in a LiPb blanket are the two isotopes 203Hg and 210Po with half lives of 46 and 138 days respectively. Both the 203Hg (g and b emitter) and 210Po (a emitter) are highly volatile materials. Polonium production can be controlled by actively limiting the concentration of the bismuth in the initial lead inventory and the bismuth that is created during transmutation of lead. Hesham assumed that 10% of the polonium and 30% of the mercury produced in the LiPb coolant would be mobilized during an accident. A leak rate of 1% per day and a containment factor of 99% was assumed. Still, the off-site doses caused by mobilization of the radioactive inventory indicate a value significantly less than that required for an emergency evacuation plan for the surrounding community. Polonium contributes approximately two-thirds of the overall off-site dose.
Loss Of Coolant Accident Analyses -Elsayed Mogahed informed the group that he had reworked his LOCA analyses to conform with the current Case J configuration, i.e., 20-cm-thick inboard shield comprised of 80% ferritic steel and 20% helium, 10-cm-thick helium manifold, 80-cm radius copper CP (24 m in length and 70% copper), and 4 LiPb cells in the outboard blanket. The 2-D analysis is conducted on a cross-sectional view at the reactor midplane. In addition to the analysis of the horizontal plane, it was suggested that a 2-D analysis be conducted in the vertical plane of the CP and the blanket to better understand how the heat is conducted down the CP and blanket.
Elsayed suggested draining the LiPb since there is a lot of heat in this coolant; but the group thought the fluid should be retained as a worst case, and it would also be representative of a loss of flow accident. It was also suggested that the maximum coolant temperature should not be used, rather the average temperature should be used as the coolant temperatures would quickly equilibrate.
Some level of conduction is required to transmit heat from the back of the hot shields to the TF coil, both inboard and outboard. Introducing a gas would accomplish this, but it was felt that this might be an active safety system. So someone should evaluate the heat leaks across these boundaries during normal operation.
Blanket Material Issues - SiC and FS - Mike Billone summarized the blanket design issues for SiC. Most of the material is being reported in his stand-alone memo on this subject. As mentioned earlier, the tritium solubility in SiC is now viewed as acceptable, although more testing is recommended. Diffusivity data would suggest it would take "forever" to diffuse through a 10 mm layer of SiC. The ability to serve as a thermal insulator is thought to be achievable with a low density SiC surface layer or SiC foam with a given porosity. Fabricating with hot-pressing is probably not feasible, but sintering with impurities may be possible, especially if reduced thermal conductivity is desirable. Chemical vapor deposition is probably not desirable in this application.
Chemical compatibility of LiPb with SiC is likely, based upon previous tests; but long-term testing is needed, especially with flowing LiPb.
GA Plasma Stability Results - Bob Miller reviewed his stability results that concentrated on ballooning stability with some supporting results on kink stability.
b bp = 25 ((1+ k2/2)( bn /100)2
A lower bootstrap fraction will require a lower bp, so one question is, can b increase by the same amount, i.e., will bN remain fixed? Bob found the answer is yes for auxiliary current drive peaked near the edge of plasma and no for peaking near the magnetic axis. :
PPPL Plasma Stability Results - Jon Menard outlined the key equilibrium and stability considerations; namely, bootstrap current fraction ~ 1.0 and as high a beta as possible.
To determine if a higher beta value can be achieved with these same basic plasma conditions, Jon took a GA pressure profile and applied it to the previous plasma of A = 1.6 and d = 0.57 and found that with a k = 3.4, a ballooning stable plasma with a bN of 8.3 and a b of 59% is possible. However, kink stability can only be achieved with a stabilizing wall at half of the previous distance, namely bwall/a = 1.1. He further analyzed other cases with the wall at the 1.2 position and determined that the baseline case is well optimized for the assumed constraints. But increased triangularity (d) improves kink stability and has a significant impact on beta when the stabilizing wall, bwall/a, is set at 1.2.
Jon thought that adding a "squareness" factor might be helpful and provide the same effect as adding edge current drive.
Vertical Stability and Equilibrium Analyses - Chuck Kessel reviewed his free-boundary solutions for the reference plasma parameters, Case J, as well as solutions with higher triangularity and lower elongations. These solutions started with 15 coils on a rectangular shape with the innermost coils inside the TF coil and the outer coil outside the TF coil. These initial solutions were iterated to yield acceptable solutions with 3, 4, 5, and 6 PF coils.
Even though the ARIES-ST designs are at a relatively low aspect ratio, which would nominally have improved vertical stability conditions, our design envelope has large plasma elongations that make vertical stability difficult. The inherent difficulty for the TSC code to solve these cases has prompted the use of the TEQ code with plasmas as close as possible to our reference case. Vertical stability cases have been analyzed using the following conditions:
These cases were evaluated for both a constant geometry plasma and a scaled plasma to obtain a constant fusion power. Impressive improvements were noted for each incremental step. All of these parameters have been adopted as much as possible within the plasma and design constraints of the existing approach.Current Drive Options and Decisions - TK Mau noted that an on-axis seed current of 0.46 MA is needed for the strawman case (Ip'/Ip ~ 99%). The following options are being considered:
[See below for final recommendations which differed from those offered during the meeting.]
A current drive system is also needed for the off-axis region near the outboard portion of the plasma. The options being considered are:
[See below for final recommendations which differed from those offered during the meeting.]
[Based upon the discussions in the meeting, a revised set of recommendations have been formulated by TK Mau as shown below.]
Steve noted several parametric changes that would enhance the beta performance, but might be difficult to implement. These were:
Steve suggested the project should consider a new baseline with the plasma elongation reduced to k = 3.0 and a d = 0.64 that results in a beta of 0.40, which would exhibit improved vertical instability. A more aggressive case proposed by GA had the k = 3.4 and a d = 0.70 that results in a beta of 0.55 to 0.70. Steve did not think this more aggressive case could be justified, but it was worthwhile to see if the new baseline could be improved toward the region of parameter space that contains the more aggressive case.
Les Waganer reviewed his progress in researching the needs for liquid fuels and the processes for producing hydrogen or synfuel using fusion energy. John Sheffield of ORNL is planning to initiate a new study of a large fusion plant that may be used to help produce hydrogen. Les Waganer will be supporting this study.
Les presented the DOE Energy Information Administration projections of petroleum resources and the need for energy over the next twenty years (to 2020). A contrary view was also shown from Colin Campbell and Jean Laherrere as reported in Scientific American, March 1998, and is shared by others. Their view is that the new oil discoveries are not keeping pace with the ever increasing world-wide energy demands. This will lead to a peak energy production occurring sometime around 2005 or shortly thereafter. Improved oil recovery, processing of oil sands, and conversion from natural gas will help delay the onset of the peak in petroleum production (and ultimately liquid fuel shortages). In order to maintain the anticipated rate of increase in industrial output and overall standard of living for the world, ultimately all the renewable energy sources must be fully utilized, including the fusion option.
Blanket and Shield System
Shield and Neutronics Analyses
Safety and LOCA Analyses
Vertical Stability and Control
Stability Studies to Achieve Higher Beta Values
Power Split Between Radiation and Heat Flux
Current Drive and Heating Requirements
To elicit some feedback from the ARIES team, Farrokh Najmabadi led a discussion of possible study topics for next year. These study topics must be of timely benefit for the fusion community and be detailed enough to warrant the capabilities of the ARIES team. The pros and cons of each topic were discussed to form a basis of selection.
Administrative1:30 Discussion of Agenda...............................................Waganer 1:35 Project Status and Plans.......................................... Najmabadi 1:50 DOE Perspective/Update.............................................Dove
Systems Studies2:10 ARIES-ST Systems Studies...........................................Ron Miller - Discussion ofCase J Strawman - Impact of Design Choices - Preparations for Final Strawman
Spherical Tokamak Engineering Session3:15 Design Integration Status..........................................Tillack/Wang 3:45 Summary of Morning CP Meeting + CP/TF Design Assessment...........................................Reiersen - Selection of CP/TF coil conductor material - Selection of cooling approach - Choice of Straight vs Flared Centerposts - Refinement of TF coils and Vacuum Vessel design - Changes to Strawman Design? 4:30 PF coil design (using Case J) and High-Temp S/C CP Option...........Bromberg 5:00 Extra General Discussion Time 5:30 Adjourn
ARIES Project Meeting Agenda
Tuesday, 5 May 1998
University of California, San Diego
Spherical Tokamak Engineering Session (Continued)8:30 Selected Power Core Configuration/Maintenance Approach......................Wang - Show coordinated CAD drawing from UCSD/PPPL (TF coils, busbars, VV,...) - Show new structural support approach - Review maintenance approach 9:15 Summary of Selected Blanket Design.........................................Sze 9:45 Extra Discussion Period 10:00 Break 10:15 Breeding and Shielding Issues and Selected Designs......................El-Guebaly 10:45 Safety Analyses...........................................................Khater 11:15 LOCA/LOFA Analyses.........................................................Mogahed 11:45 Extra Discussion Period.....................................................All 12:00 Lunch 1:00 Blanket Materials Issues: SiC and FS......................................Billone 1:30 Testing Results of a Tungsten Rod PFC Surface..............................Waganer 1:45 Wrapup of the Design Approach for ARIES-ST.................................Tillack 2:15 Extra Engineering Discussion Time.............................................All 2:30 Break
Spherical Tokamak Physics Session2:45 Plasma Stability Results ....................................................Jardin 3:30 Parametric Plasma Stability/Performance Results.............................Stambaugh 4:00 Extra Discussion Period ........................................................All 4:15 Recommended Startup, Heating, and Current Drive Options.........................Mau - Show location and volume of proposed antennas and power required 4:45 Extra Discussion Period .........................................................All 5:00 Summary of Recommended Plasma Parameters and Systems.........................Jardin 5:30 Adjourn for Project Dinner
ARIES Project Meeting Agenda
Tuesday, 5 May 1998
University of California, San Diego
Alternate Applications Study8:30 Review of Progress in Alternate Applications Study.....................Waganer
Overview of ARIES-ST(9:35 to 11:00) Summary of Physics Plans.......................................................Jardin Summary of Engineering Plans (may do at the end of Tuesday).....................Tillack Summary of Systems Plans.........................................................Miller Project Next Actions to Prepare for Design Completion...........................Najmabadi ANS meeting plans and papers?...................................................Najmabadi Next Meeting Arrangements (6-7 August at ANL)...................................Najmabadi/Sze 11:00 Adjourn