The main purpose of the meeting was to review the physics and engineering analyses in preparation for the selection of the parameter space for the near-final strawman design. This "near-final" strawman is to be issued in late November 1997, approximately 6 weeks ahead of the next ARIES project meeting, scheduled for 8-9 January at San Diego. The next project conference call is scheduled for 22 October.
A set of plasma parameters were presented over a range of aspect ratios of 1.2, 1.4, 1.6 and 1.8. These sets should be refined to agree with chosen engineering options and constraints. A case with finite edge density should also be analyzed with respect to both physics and engineering considerations. The engineering aspects of the subsystems were discussed, with key issues being highlighted in each area. All project activities are being focused to choose compatible design options to meet the overall performance, maintenance, and economic goals. The issuance of the November strawman should allow sufficient time to develop detailed system and component performance data for the January meeting.
Rob Goldston, the Director of PPPL, praised the ARIES past and present efforts to establish technical direction for the entire fusion community. He encouraged the group to next examine the Field Reversed Configuration (FRC) or the high-beta Stellarator as a commercial power plant.
Bill Dove noted that a team from the University of Washington, the University of Wisconsin, and the University of Illinois, led by John Santarius (University of Wisconsin) and assisted by Loren Steinhauer (University of Washington) and George Miley (University of Illinois), is assessing the engineering issues of an FRC design. In another related, but separate study, Loren Steinhauer and Steve Jardin will be assessing other aspects of the FRC approach. Bill mentioned that there are several new Spherical Tokamak (ST) devices that are being designed and built around the world, reflecting increased interest in this concept. He noted that there has been nothing released about the 1998 budget, but later in the meeting Rob Goldston informed the group that the approved ER budget is to be $232M.
Farrokh Najmabadi told the group that the ARIES-ST project is progressing on schedule. The parameter space and subsystem options are being examined for key issues and general performance capability in an ST power plant. Plasma profiles are being defined to yield a high bootstrap fraction (~99%), but some current drive power may be required to provide an adequate design margin (for example, the current drive may need to drive about 5% of plasma current). The centerpost design is not yet consistent between magnet designers, systems code, and maintenance concepts. The divertor group is struggling to define concepts that will accommodate the required heat load and allowable space envelope. Hopefully, a consistent design can be completed and documented by Spring 1998.
To acquaint the ARIES team with the latest thinking of two promising confinement concepts, Sam Cohen andAlan Reiman presented data on the advanced FRC and high beta stellarator concepts, respectively, Sam thought economics should not be used as a criteria to judge the merits of a confinement concept. Instead, a project should first concentrate on engineering issues, with subsequent direction to the physics team to provide a final solution. Since many engineering problems are neutron related, advanced fuels which produce "neutronless" or reduced neutron reactions should be examined in more detail. Sam showed a sketch of an advance-fueled FRC which extracted 80% of the fusion energy thermally and 20% kinetically through a turbine-generator set.
Alan Reiman described the High Beta Stellarator with quasi-axisymmetric (QA) fields as a nearer term device with attractive features of a higher beta at a relatively low aspect ratio and possibly lower cost. He applied QA fields to the ARIES-RS plasma configuration to obtain a 2-period QA stellarator. Ballooning modes may be a problem for the QA stellarator.
Gary Voss from Culham attended our meeting to observe our design status and evolution process. We asked him to update our team on Culham's programs and current status. He is responsible for the MAST coils. The MAST experiment is an ST design. He showed a cross section of the Culham ST power plant design. Two TF coil design concepts are being considered: water-cooled GlidCop coils and cryogenically-cooled aluminum coils. The gas-cooled first wall is maintained from the inside of the machine using removable panels which have a 2-year lifetime. With a 10 to 20-cm thick inboard shield, the shielded center post is predicted to have a 2-year lifetime. The breeder material is either LiH or LiO2.
Equilibrium Results - Jon Menard presented the most recent plasma ballooning stability results over the range of aspect ratio of 1.25 to 2.0 and k (elongation) =1.6 to 2.8 @ d=0.45 (lowest useful triangularity). The kink stability limits were studied as a function of k and d @ A=1.4 and fixed bN. The beta limit for A=1.4, 1.6, and 1.8 @ d =0.57 were simultaneously optimized for kink and ballooning stability. The effects of lower triangularity on kink and ballooning stability at fixed k at A=1.6 were studied as well as keeping the stabilizing wall position fixed. All equilibrium cases were developed with a bootstrap current fraction of approximately 99%. It was suggested that it might be better to have some current drive (CD) capability with a reduced bootstrap fraction.
Stability Results - Chuck Kessel explained the difficulty of obtaining free-boundary equilibrium conditions with a low aspect ratio. He modified the code equation to allow better convergence at low A. He illustrated the results at A=1.4, 1.6, and 1.8 to the group. A large number of PF coils were used but these should be approximated with a reduced number set. Since presently there are no inboard PF coils, the inboard plasma radius is not directly controlled; hence an inner coil may be required. Chuck also illustrated three possible poloidal flux geometries outside the plasma that affect the heat deposition-limited plasma with no X-point, limited plasma with an X-point, and a diverted plasma boundary attached to an X-point.
Current Drive and Startup - T.K. Mau presented the on-axis seed current drive system results for an aspect ratio of 1.6 using an updated version of CURRAY. He noted there remain some discrepancies in the results of the JSOLVER and CURRAY codes for on-axis current drive requirements, which indicate a need for added data points from C. Kessel. Plasma edge density is very important to the bootstrap current drive requirement. He determinedthat the seed current should be 0.29 MA on axis. This value is sensitive to lower electron plasma temperatures. For the on-axis seed current drive, both LFFW (low frequency fast wave) and ICRF (ion cyclotron resonance frequency) fast wave frequency systems were investigated but the LFFW is the only theoretically plausible RF technique that drives currents near the magnetic axis. HHFW (high-harmonic fast wave) does not penetrate because of strong electron and/or ion damping, and the ICRF fast wave system suffers strong electron and alpha/ion damping. Off-axis current drive is best accomplished with HHFW or ECH (Electron Cyclotron Heating). Target plasma formation is the first phase of plasma startup (up to 200 kA). The second phase is to ramp up the current using bootstrap current overdrive. The CURRAY code now has the capability to analyze these equilibria and perform the analysis for the HHFW heating.GA Physics Study Results - No results were presented. Physics Summary - Steve Jardin highlighted the following data.
Ron Miller reported that he had completed a strawman data set and posted the results on an accessible web (http://aries.ucsd.edu/miller/). Several key results from this strawman were presented at the meeting. Now that the A=1.8 physics results are available, this data set will be evaluated with the systems code and all prior data sets will be updated to agree with new engineering design options and constraints. But the new strawman is to have an aspect ratio of 1.6 and is to be available by late November 1997.
There was concern that the systems code representation of the centerpost and divertor regions were not consistent with the engineering design approaches and the maintenance configuration. Ron agreed to quantify the benefits of a shaped CP as opposed to a straight cylinder. His code is presently only examining the general parameter spaces. When baseline engineering and physics designs are adopted, his code will be revised to have a higher level of fidelity. The present coil sets and plasma field surfaces are well matched with the physics and engineering parameters. Plots of the poloidal varying neutron wall load are being provided to L. El-Guebaly for neutronic calculations.
To date, narrowing the design approach to these three options has been based on the engineering feasibility; however, the final selection between these three options will require a determination of the capital cost of each option along with the expected time and operational cost to conduct both scheduled and unscheduled maintenance.
Magnet Systems Design Studies - Wayne Reiersen reviewed the objectives and progress on the center post and TF coil design study. The magnet system design analysis code has been updated to include the current CP and magnet configuration at A=1.5 and a 30-cm-thick inboard shield. This code has also been benchmarked against F. Dahlgren's 3-D thermal-hydraulic code. Results of the code were presented for consideration. Sliding electrical contacts have been incorporated into the upper TF coil legs near the CP. Stress calculations have been conducted with this configuration. Design analyses have been conducted for the CP with both brittle copper (low-temperature water cooling) and ductile copper (moderate temperature water) and it was concluded that it would be better to use dispersion-strengthened copper (GlidCop AL-15) with an inlet water temperature in the range of 150-180oC. Cost algorithms for the power supplies were estimated to be in the range of $100/kVA.
Evaluation of ST Blanket Options - Dai-Kai Sze noted that a lithium, self-cooled blanket is feasible and does not need an insulating coating. However, the concern about using a water-cooled centerpost in the presence of a lithium-cooled blanket prompted Dai-Kai to concentrate on developing the design of the helium-cooled ferritic steel and SiC lithium-lead breeding blanket (He/FS/SiC/LiPb). He proposed recommendations for several issues that had arisen over the past meetings.
It was recommended that the He/FS/SiC/LiPb first wall and blanket design be incorporated in the next strawman and design assessment. Helium coolant first cools the separate, thin FS first wall and then enters the center of the blanket nearest the plasma and flows past the SiC plates that contain the LiPb breeding material and down into the SiC block for superheating the helium coolant. The first wall will not be smooth, rather will be a series of adjacent cylinders (as shown in the following sketch). At present, there is no separable zones that are matched to predicted operational lifetimes.
Possible Liquid Target Divertor Options - Mark Tillack stated the positive features of the liquid divertor options that motivated an assessment for the ARIES-ST design, namely, unlimited lifetime with no neutron or transient damage, increased operational temperature limits, increased operational lifetime, reduced plumbing connections, and possible active pumping. This idea has been promoted previously in magnetic confinement and is one of the leading plasma facing component design options for the inertial fusion studies. Lithium, gallium, or possibly a eutectic of lithium and gallium are leading material contenders. Liquid films seem to be better at handling high surface heat loads as opposed to droplets. Mark mentioned that he is a participant in the advanced liquid-protected surface (ALPS) planning group to examine such a concept and hopes to adopt some of their ideas in our design as a benchmark evaluation. Present guidance is that 10 to 15% of the divertor power will be present on the inboard divertor target area and 85 to 90% of the power will go to the outboard divertor target area.
In addition to the liquid target, Mark will investigate the radiative mantle, radiation of ~75% of the divertor power to the entire plasma facing wall, and the vanadium divertor surface (maximum heat flux limited to approximately 5 MW/m2).
Stress Analysis of TF Coils for a Low Aspect Ratio Tokamak- Leslie Bromberg presented the stress and strain analysis of a TF coil set for a previous low aspect ratio strawman. Unfortunately, the coils are not demountable and did not contain a sliding contact joint to reduce the stresses in the centerpost. The stresses analyzed were acceptable, even in the absence of joints (joints may be required for other configuration reasons). The case analyzed seemed to have unusually low stresses on the centerpost. Leslie attributed this effect to the restraint of the PF coils, which were structurally attached to the outer legs in his design approach. The results indicate that the vertical field coils need to be included in the inplane analysis, since they dominate the results. The PF coil acts as an electromagnetic press in this case. Although Leslie did not analyze the current TF coil design, but the results are indicative of the general class of low aspect ratio TF coil sets.
Algorithms for the centering force were developed for Ron Miller's system code. A vertical loads algorithm, however, was not developed because of the strong PF influence discovered during the analysis. Leslie agreed to redo his calculations to eliminate any constraints between coil sets. Leslie did not design the PF coils, rather he used the design information in the latest strawman data set. He recommended using superconducting PF coils, probably NbTi superconductors.
Tritium Breeding Issues and Radial Builds - Laila El-Guebaly presented results of tritium breeding analyses and suggested radial builds for the ST fusion power core. She also concentrated on the He/FS/SiC/LiPb blanket design. She will coordinate with Sze for the latest information regarding this LiPb blanketconcept. For a case with an aspect ratio of 1.6, the percent of neutrons absorbed in the inboard first wall region is 9% and 84% in the outboard region with the remainder to the divertor. The tritium breeding ratio is greater than 1.1 if water is not used in the inboard shield. (Water in the copper centerpost is acceptable.) She is presently predicting the 3-D results based on a 1-D analysis with extrapolation. After the design is nearly final, a full 3-D analysis will be conducted. At present, simplifying assumptions are being used for penetrations, but soon T.K. Mau can provide more specific RF system area and material data.
For the He/FS/SiC/LiPb blanket design, the thickness of the SiC insulation for the first cell principally determines the achievable breeding ratio. The nominal He/FS/SiC/LiPb blanket design used 1-cm-thick SiC layer inside the ferritic steel (FS) cells; but much better breeding performance is obtained with eliminating or reducing the thickness of the SiC in the first cell. She emphasized the need to optimize the material choices at the front of the blanket. For a peak wall loading of 10 MW/ m2, the 1.3 FPY lifetime of the first wall and blanket infers 30 replacements in the 40 FPY power core lifetime. Laila will send the composition of the DS GlidCop AL-15 to Tillack and Wiffen.
Safety - L. El-Guebaly presented the LOCA safety analysis of E. Mogahed, H. Khater, and L. El-Guebaly which was updated to include the ultimate heat sink of the TF busbar. With the large thermal bus bar heat sink, the center post temperature during LOCA will not exceed 200oC. The maximum FW temperature during LOCA is dependent upon the average inboard neutron wall load and the initial temperature of the inboard shield. The heat transfer from the shield to the copper CP is limited by radiation across the gap between the shield and CP. (No conduction is assumed.) It was recommended that Reiersen et al. define a method to passively conduct heat across that thermal interface. It was affirmed that the baseline assumption for the LOCA is a complete loss of coolant to the CP as the worst case scenario. With the present design (radiation only) and the maximum neutron wall load of 5 MW/ m2, a complete LOCA with no passive cooling loop will cause the FS inboard shield temperature to be between 850-900oC.
Design Rules - L. El-Guebaly presented the brittle material design rules prepared by J. Blanchard and J. Crowell. For a brittle material, a thin first wall can tolerate double the heat flux of an integrated first wall and blanket. Creep lifetimes of martensitic steels are typically 5 years or over. Thermal creep in ferritic steels requires operations at less than or equal to 550oC, whereas irradiation creep is minimal. Thermal creep in vanadium is minimal up to and including 600oC, but no data is available above that level. Irradiation creep in vanadium is a design concern.
Les Waganer updated the team on the alternate fusion product study plan to collect and categorize fusion applications and formulate a methodology to assess the attractiveness of the alternates. The data sets were revised to include comments from the project team at the last meeting and inputs from several advocates. These revised data results were presented to Anne Davies and the Office of Fusion Energy Sciences in early August as well as at the Fusion Power Associates meeting in early September. It was recommended that the methodology and results be sent for critique to additional product advocates.
Les also presented another method to analyze engineering designs, using a parameter space of design difficulty and resources. The case studies used to validate the approach ranged from the simple and inexpensive to the complex and costly. The general approach and validation methodology were similar to the method proposed for fusion products, but the metrics chosen w ere somewhat different.
This is the time to document the findings and results from the engineering evaluations and analyses. For the strawman design, an aspect ratio of 1.6 will be adopted, no breeding is assumed on the inboard regions, and 50% core radiation is adopted. Specific action items are noted below.
8:00 Coffee and Socializing...............................................PPPL hosts
Administrative8:30 Welcome and Administrative Details......................................Jardin 8:35 Welcome to Princeton....................................................Goldston 8:45 Discussion of Agenda...................................................Waganer 8:45 DOE Update .......................................................Dove 9:00 Program Overview and Plans.............................................Najmabadi
Invited PPPL Talks9:20 A Fusion Concept Based on the FRC and Advanced Fuels that minimizes plasma/surface interaction....................... Sam Cohen 9:40 New Developments in High-Beta Stellarators with good confinement through quasi-symmetry................................ Alan Reiman 10:00 Break
Spherical Tokamak Physics Session10:15 MHD stability limits as a function of Aspect Ratio and Triangularity at Low Aspect Ratio......................................Jon Menard 10:45 Free Boundary Equilibrium Considerations at Low-A......................Kessel - Equilibrium Results at 1.8? 11:15 GA Physics Study Results............................................ Stambaugh/Miller? 11:45 (Makeup time) 12:00 Lunch to 1:00 1:00 Current Drive and Startup at A=1.6....................................... Mau 1:30 Discussion of ARIES-ST Physics and Engineering Interaction...............Jardin 2:15 Break
Spherical Tokamak Engineering Session2:30 Assessment of Design Parameter Space (A=1.2 to 1.6).................Ron Miller Status of Design Assessments...... (Options, Major Issues, Recommendations) 3:00 Power Core Configuration/Maintenance Option Assessment........Wang/Najmabadi 4:15 First Wall, Blanket, and Power Conversion Assessment..................Sze 5:00 Assessment of Liquid Target Divertor Options...........................Tillack 5:30 Adjourn for Project Dinner.................................Project
8:00 Coffee and Socializing......................PPPL hosts
Spherical Tokamak Engineering Session (Continued)8:30 TF Coil Structural Analysis + Busbar and Power Supply Info.............Bromberg 9:00 Breeding Issues and Recommended Radial Build...........................El-Guebaly 9:45 Effect of Coil Bus on Center Post LOCA Temperature......................Mogahed 10:15 Break 10:30 Update on Structural Material Design Code Recommendations...............Blanchard 11:00 Summary and Selection of Engineering Design Options......................Tillack 12:00 Lunch to 1:00
General1:00 Assessment of Alternate Fusion Applications............................ Waganer 1:30 Physics Summary and Direction for Next 6 Months.........................Jardin 2:00 Engineering Summary and Direction for Next 6 Months.....................Tillack 2:30 Break 2:45 General Discussion Period (Slack Period) 3:30 Action Items, Plans, Report Status........................................Najmabadi 4:00 Adjourn