The meeting for the ARIES Advanced Concept Studies program was held at University of Wisconsin at Madison in conjunction with an ARIES-sponsored town meeting on designing with brittle materials. The agenda for the town meeting is appended to the attached project agenda. Separate meeting minutes will document the key actions and results from the town meeting.
The main purpose of the ARIES project meeting was to assess the physics and engineering options and approaches for the commercial LAR/ST design. The intent is to establish an initial baseline design within six weeks so that all analyses and design work and trade studies have a common basis for comparison. This will allow more focused results. The next meeting at PPPL is currently scheduled for 25-26 September. The preliminary results for the alternate products study were also reviewed in preparation for a DOE briefing in July.
Jerry Kulcinski welcomed the ARIES team to the University of Wisconsin. Bill Dove noted that the 1998 budget authorization has been completed and is working its way through Congress. Funding for the current Advanced Concept Studies Program is slated to be around $2M. The 1999 budget is expected to be at a similar level. Bill mentioned that DOE is looking for additional support to stress the importance of fusion over a broader range of applications, hence the timely briefing on alternate products. He also secured a briefing to DOE on the ARIES-RS results on 2 July. A new approach starting in FY99 is to have a plan stressing the project performance, namely specifying the deliverables, such as reports or analyses. So he is encouraged that some ARIES reports are soon to be published. He is also investigating how to publicize that the ARIES reports are on the public web servers.
Farrokh Najmabadi discussed the Advanced Concept Studies Program schedule and explained that the current examination of the LAR/ST design is really intended to be a two-year effort. The June 1997 meeting examined design concepts and performance parameters for the major LAR/ST subsystem options. The intent to select subsystem options within six weeks, which would allow establishment of a baseline strawman and assessment of a cohesive, integrated system design and reference case by the end of 1997. Then the LAR/ST concept can be evaluated on a comparative basis as a commercial electrical generating power plant. Interaction between team members is becoming more important in this phase; hence, the team should increase the use of project conference calls, perhaps monthly.
High-harmonic fast wave (HHFW) as a profile control technique was shown to be less effective in higher-beta ST devices such as ARIES-ST at A=1.4 where power deposition is now limited to y >0.75.
Bootstrap current ramp-up, as proposed by Nevins, appears attractive for ST devices. Full current ramp-up in a fusion power plant using this technique plus non-inductive on-axis drive remains to be demonstrated by simulations. Formation of the target plasma for this process is critical. Coaxial helicity injection is problematical in view of the physics uncertainties, lack of active profile control, and the need of insulating rings close to the plasma. A scenario involving both ECH and RF power for heating and current drive should be considered.
Ron Miller reported that he is now officially in charge of the ARIES systems code. He has completed incorporation of the ST physics model for the A=1.25 and A=1.4, but the A=1.6 requires inputs from C. Kessel, who is unavailable for the immediate future. Engineering inputs are being added as they become available. The strawman engineering options are to be decided in six weeks at which time the modeling algorithms should be added to the code [Action Item]. This will allow the new strawman data to be generated in time to be analyzed by the subsystem design leaders for the September meeting.
Ron reviewed trends for fuel and electricity prices in the U.S. Even though aggressive requirements were established for ARIES-RS, it seems the fusion goals may not be in a competitive range according to the available information. More aggressive goals must be set in order to have a competitive product.
Ron showed the systems code component geometry and noted that it is not yet consistent with the proposed coils, blanket/shield, and configuration/maintenance geometries. Specific design options need to be added (see first paragraph in this section) for an integrated assessment and optimization.
Les Waganer reviewed the alternate fusion product study plan to collect and categorize fusion applications and formulate a methodology to assess the attractiveness of the alternates. The chart of the potential product lines was reviewed. The list of product descriptions was distributed at the meeting for review. The decision analysis methodology was examined by the ARIES team. Attributes were identified along with comparative examples for assignment of quantitative values. The attributes were assigned weighting values according to the perceived value of the attribute. A set of major national projects were evaluated with the methodology and the results correlated well.
Les then presented the preliminary results of applying the methodology to the proposed fusion products. The end results seemed to be insensitive to the assigned weighting values. The team recommended testing several other weighting approaches. Several assigned attribute values for various applications were discussed. The team generally agreed with the assigned values, but several members volunteered to provide alternate weighting and attribute values for comparison. Les will present the preliminary results to DOE in early July.
Other recommendations include encasing each superconducting PF coil in a separate cryostat rather than a large common enclosure. All designs offered the option to remove the central post separately from the remainder of the replaceable power core. A curved centerpost would severely constrain the separate removal of the centerpost. The need to define the vacuum vessel geometry was stressed so that the maintenance plan can be more fully developed.3-D Neutron Wall Loading and Blanket Neutronics - Laila El-Guebaly presented the 3-D neutron wall loading (NWL) results (shown at the right), which indicated that the inboard NWL is nearly constant from the center region to near the divertor. On the outboard region, the NWL peaks at the center plane, but the region near the divertor is less than one-fifth the peak value. The NWL in the divertor region is very low. For the outboard blanket, the neutron coverage fraction is 90% for A=1.25 and 84% for A=1.6.
There are a few discrepancies between the MCNP and the ARIES System Code results (in the divertor region) that need to be resolved before the strawman is released.
Laila presented a first wall plot that compared the ARIES-RS geometry to the LAR/ST geometry that emphasized that the volume and surface area of the LAR/ST device is substantially larger than the reversed shear device.
Since the geometry of the LAR/ST device has not been optimized and selected, Laila examined the A=1.25 strawman with 90% enriched LiPb breeder to determine the ability to adequately breed tritium (TBR greater or equal to 1.1) with only an outboard breeding blanket. With ferritic steel structure and a SiC insulator, the TBR is marginal, but elimination of the SiC raises the overall TBR to 1.24. The outboard neutron coverage fraction (and breeding) decreases with increasing aspect ratio. Addition of a kink stability shell also has a negative influence on the TBR. Replacing the LiPb with Li would degrade the TBR when the same structural materials (FS and SiC) are used. For the same blanket thickness, a Li/V blanket gives higher TBR than a LiPb/FS/SiC blanket. An analysis of the free-surface Li2O pebble breeder indicates adequate tritium breeding. Depending on the type of breeder, ST machines with A > 1.25 may need to breed on the inboard and/or divertor or use a beryllium multiplier for an adequate TBR. The overall energy multiplication of the system is approximately 1.1.
Laila continues to size the first wall, blanket, and centerpost (CP) shielding so that all components will reach their end of life at the same time, which eases maintenance operations and enhances the replacement costs. This effect has not been fully integrated with the CP design to date.Embrittled Material Design Issues for Fusion Devices - Jake Blanchard explained that the ASME and RCC-MR codes were formulated mainly for pressure vessels designed with ductile materials (limit average primary membrane stress < Sm, limit total primary stress and primary + secondary to < 3Sm). ITER is modifying and adding to the design rules to include embrittled materials. This includes modified primary membrane stress, new limit on primary + secondary membrane stress, and modified limit on primary + secondary stress. Irradiation creep and swelling are also addressed to maintain structural adequacy. To assess the effect of these new ITER rules, Jeff Crowell applied them to the DS Cu CP and the V-4Cr-4Ti first wall designs. The irradiation in the outer region of the Cu CP causes the material to embrittle very quickly, with a corresponding low allowable stress. If the component operates at a higher temperature, the material softens and the ultimate and yield strengths decrease to a fraction of the room temperature properties. The CP should be well shielded and cooled for best strength properties.
The V-4Cr-4Ti first wall design, used as an example, was attached to a relatively stiff blanket which caused high stresses in the first wall. This drove the design parameters to low surface heat flux wall loadings and/or thin walls. The solution is to provide more flexibility in the structure or first wall attachment. Nonetheless, the new design rules reduced the allowable surface heat flux by nearly a factor of two.
Jake will work with ITER JCT to help tailor the design rules for commercial power plant designs.ARIES-ST Centerpost Design Studies - Wayne Reiersen is evaluating design approaches for the ST CP. The basis of comparison was the average of A=1.4 and A=1.6. Some design options are:
The results of a trade study of several CP options operating with different materials and temperatures were discussed. Heat recovery and cryogenic refrigerator efficiency were included. The group suggested the cryo pumping efficiencies were too pessimistic. A cryo-resistive OFHC Cu CP did not offer any significant advantages with the data used (I. Sviatoslavsky suggested using 40% of Carnot instead of 20% of Carnot. The 20% number would be reasonable for lower temperature refrigerators. The LN2 calculations will be redone with Igor's numbers. Igor also suggested that the HP aluminum option would look better in the 30-50K range rather than the 10-20K range. That option will also be explored.). The attached sketch of the PPPL-proposed configuration was used in the CP evaluation, but it has not been integrated with the UCSD maintenance efforts.
A multi-mega amp power supply scheme has been developed that looks reasonable.ARIES-LAR Center Post Design and Performance - Fred Dalghren compared design parameters for center posts with shields of 30, 20, 10, and 0 cm thicknesses. The conductor area in the PPPL design code increases as the shield decreases. Voltage feedback is included in the analysis to hold the current constant over the life of the CP. The least power is consumed in the unshielded CP because of the extra copper in this design. But again this probably is not an acceptable design because of the radioactive waste and strength concerns in an unshielded CP.
The wedge design approach for the CP has been redesigned to incorporate coolant channels. A thermal stress analysis was conducted to determine if the design is feasible. Some modification and relocation of the channels may be recommended to minimize the thermal stresses. The team recommended that the center hole be more effectively used to increase the copper content of the CP. Fred also presented a sliding electrical contact system for the upper and lower legs of the TF coils. .Comparison of 30K Super Conductor Busbar and Power Supply Systems - Leslie Bromberg examined the helium-cooled superconductor (NbTi) lead systems and concluded that they require excessive coolant and are too expensive. Likewise, the low critical temperature super conductors are not feasible.
Evaluation of ST Blanket Options - Dai-Kai Sze explained that the ST power plant will most likely have a water-cooled center column. However, there are safety concerns between the water and first wall and blanket materials (such as lithium, lithium-lead, and beryllium). Hence, the blanket material must satisfy certain safety issues. Moreover, it must have the potential for high neutron wall loading > 5 MW/m2. Because the CP causes a large recirculating power fraction, the blanket must operate at a high temperature to allow for a very efficient thermal conversion system. And since there will be no inboard breeding blanket, the outboard blanket must breed all the tritium required for operation.
The lithium/vanadium blanket is simple, can operate at high temperature, has a high tritium breeding ratio, can handle high neutron wall loads, and has low activation materials. The ST geometry may allow low pumping power operation without an insulating coating. But the interaction with the CP water coolant remains a serious safety issue.
The He-cooled/Li/V design is more complicated that the Li/V design. The high pressure helium coolant is used with a closed-cycle gas turbine in order to have a high efficiency conversion system. The wall loading and tritium breeding are adequate but inferior to the Li/V design.
The He/FS/Solid Breeder/Be/SiC design concept uses high pressure helium. SiC is used to thermally isolate the blanket to allow it to run to higher temperatures while keeping the FS structure within allowable temperature limits. This results in a complicated breeder design. The neutron wall loading may be limited by the ability to cool the first wall and FS structure.
The He/83Pb17Li dual coolant design is based on the EC demo design approach and the He/FS/Solid Breeder/SiC design. FS is again the structural material and is cooled with the low temperature He. The 83Pb17Li is both the high temperature coolant and breeding material. SiC is used for thermal and electrical isolation between PbLi and FS. This system should be safer than the others. The thermal efficiency is on par with the others. But the system is complicated, and the first wall heat transfer is limited.
The newest option is a free surface LioO particle system using a free gravity drop of particulates with SiC guides to locally form and control the flow of the particles. This solid breeder may be able to adequately breed without beryllium, and high neutron wall loads may be possible. Structural lifetimes may be extended. But particulate handling and heat removal are definite concerns.Safety Issues Waste Disposal Ratings - Hesham Khater noted that the beryllium reaction with water is much less severe if the material remains in solid form as compared to being in a liquid form (550 GJ vs. 16,680 GJ for a design containing 300 tonnes of Be). He presented data on a shielded CP with F82H steel, a CP shielded with ORNL steel, and an unshielded CP. An unshielded CP will not meet the Class C waste with either the NCR or the Fetter limit criteria. The F82H steel is slightly superior with the NRC limits, but the ORNL steel is superior under the Fetter groundrules. LOCA Analyses - Elsayed Mogahed reviewed his LOCA analysis basis and his results. He modified his analysis groundrules to allow conduction up/down the center post, but connected to an ultimate heat sink only through the steel jacketing at the outer radius of the CP (the remainder radiates to the ultimate heat sink). This still caused high temperatures in the CP during a LOCA. The present CP design is intimately connected to very large busbars (an ultimate heat sink) either at the top or the bottom of the CP and also is connected to the TF coil outer legs at the other end of the CP (also an ultimate heat sink.) Elsayed will redo his analysis with the new information and keep appraised of the current design via M. Tillack or W. Reiersen. Controlling the Radiological Hazards of LiPb - Layton Wittenberg explained that the major sources of a radiological hazard in a PbLi blanket are the two isotopes 203Hg and 210Po, both of which are highly volatile materials. The 210Po is considered the main safety hazard. It is produced from either 208Pb or 209Bi (impurities in Pb). By controlling the concentration of bismuth in the commercial lead (typically 500-1500 wppm) to below 10 wppm, it is felt the radiological hazard from 210Po is minimized. This cleanup would be required for the initial supplies and on a continual basis to keep the concentration of bismuth at an acceptable level. The removal of bismuth can be accomplished by cooling the liquid to 300oC to allow growth and separation of Li3 Bi crystals.
8:00 Coffee and Socializing........................... UW hosts
Administrative8:30 Welcome to the University of Wisconsin........... Kulcinski 8:40 Discussion of Agenda............................. Waganer 8:45 DOE Perspective.................................. Dove 9:00 Program Overview and Plans....................... Najmabadi
Spherical Tokamak Physics Session9:15 a) Radiative Mantle Experiments on TFTR with ........ Jardin Kr and Xe and Implications for Power Producing Plasmas b) Free Boundary Equilibrium Considerations at Low-A.... Jardin 10:00 Break 10:15 GA Physics Study Results......................... .Robert Miller - Stability Studies with Non-Zero Edge Current - Stability Studies of beta vs. aspect ratio - Startup Scenario- betaP overdrive 11:00 Startup and Current Drive Systems......................Mau - CD Options - Startup Scenarios (HHFW and Bootstrap Overdrive) - Beta Enhancement with Edge CD 11:45 Discussion Period 12:15 Lunch to 1:00
General1:00 ARIES SystemsStudies....................................Miller - Results of Strawmen Runs at A=1.4, 1.6 - Comparison of Parameters, Assumptions, and Expected Performance of Four LAR Concepts - Results of Four Centerpost Strawmen Comparisons (From Reiersen's Designs) 1:45 Assessment of Markets and Customers for Fusion Applications.....Waganer - Discuss Remainder of Products - Present Attributes for Decision Criteria Methodology - Review Examples of Past "New" Technologies - Discuss Trial Results 2:45 Break
Spherical Tokamak Engineering Session3:00 Evaluation of ST Maintenance Configurations................Najmabadi/Wang - Vertical vs. Horizontal Maintenance Approaches 3:45 Safety Analysis Results for LAR Tokamak.........Khater/Moghaed/Wittenberg - Include Axial Thermal Conductance - Compare Results to a Bare Centerpost 4:30 3-D NWL and Blanket Neutronics................................El-Guebaly 5:15 Adjourn for Project Dinner.............................Project
8:00 Coffee and Socializing........................................UW hosts 8:30 Fusion Design Guidelines for Brittle Material Usage...........Blanchard 9:00 CP Design Options and Performance Parameters.........Reiersen/El-Guebaly - Compare Four Design Options (Matls, Radial Build, Waste, Resistivity, Lifetime, Power Rqmts) - Electrical and Coolant Routing Details 10:00 Break 10:15 Copper Center Post Design Details and Performance Results....Dahlgren - Revisions to Modeling Code (Ohmic Heating, Constant Current) - Incorporation of Insulators - Increase of Copper Conductor Diameter when Shield is Thinned - Wedge Option Design Results 11:00 Discuss ST Blanket Options...................................Sze - He-Cooled FW with Li-Cooled Blanket - H20-Cooled CP with Li-Cooled FW & Blanket w/Protective Barriers - Free-Surface FW of LiO2 Pellets 12:00 Lunch to 1:00 1:00 Discuss Designs of 30K S/C Busbar and Power Supply System.....Bromberg - Compare to Copper Option 1:30 Summary of Engineering Options and Selection of Concepts......Tillack 2:15 Break
General2:30 Physics Summary and Direction for Next Six Months.............Jardin 3:00 Engineering Summary and Direction for Next Six Months.........Tillack 3:30 General Discussion Period (Slack period)......................All 4:30 Action Items, Plans, Report Status...........................Najmabadi 5:00 Adjourn
8:30 Welcome, Objectives of the meeting.............................Tillack Welcome from the University of Wisconsin............................Kulcinski 8:45 Introductory remarks from OFES.................................Wiffen 8:50 ARIES Overview.................................................Tillack 9:30 Fusion power plant design issues related to brittle materials..Blanchard 10:00 Latest thinking on blankets for low-aspect-ratio tokamaks.....Sze 10:30 Centerpost issues for spherical tokamaks......................Dahlgren 10:45 Break 11:00 Failure mechanisms & design rules with brittle structural ...Majumdar - Materials - ITER perspective 11:30 Bonded structures, ITER and Aerospace perspectives...........Davis 12:00 Lunch 1:30 Fission reactor pressure vessel insights & master curve.......Odette approach to fracture toughness assessment 2:00 Radiation embrittlement of FCC metals (Cu, austenitic SS)........................Zinkle 2:30 Radiation embrittlement of BCC metals (F/M SS, V-alloys)......Zinkle 3:00 Utilization and data base issues for SiC/SiC..................Jones 3:30 Discussion, action items, future plans...................Tillack, et al 4:30 Adjourn