ARIES Project Conference Call

2 March 1998

Documented by L. Waganer

Participants:

(DOE)	*
(UCSD)	Mau, Ron Miller, Najmabadi, Wang
(PPPL)	Kessel
(GA)	Bob Miller, Stambaugh
(UW)	El-Guebaly, Khater, Mogahed
(FPA)	*
(RPI)	Steiner
(ANL)	Sze
(Boeing)	Waganer

Administrative

The dates for the next ARIES project meeting had been set for 29 April through 1 May in San Diego; however, Bill Dove asked to have the ARIES meeting contiguous with a town hall meeting. Farrokh Najmabadi was considering delaying the project meeting a week, which posed no problems to those on the call. The meeting date will be set by 6 March. Les Waganer said that the UCSD rate at the Marriott Residence Inn had raised to $129 - Ron Stambaugh will confirm the GA rate at that hotel and advise the team of his findings. (Mark Tillack later reported that the ITER and GA rates for the Residence Inn are $99.)

Laila El-Guebaly offered to host the following project meeting at Madison, suggesting the dates of 29-31 July or during the first week of August. Instead, Dai-Kai suggested hosting it at ANL as Chicago is more accessible, has cheaper airfares, and the meeting can be completed in two full days as late flights can reach both East and West Coasts on Friday evening. The group agreed to an ANL meeting, tentatively set for 6-7 August. For directions to ANL, you may access this site . If anybody has a conflict with those dates, contact L. Waganer. You might also want to consult with the ARIES Team Calendar on the ARIES Web page.

At the end of the call, everybody agreed there were a lot of open-ended technical discussions, without convergence, involving both Physics and Engineering. Hence, it was decided that rather than have another project conference call in 3-4 weeks, it would be more beneficial for the Engineering group to have a call on 16 March (# is 314-232-9568); the Physics group would have one in the same time period (date, time, and call number to be provided.) Then the Project will have a call on 13 April (# is 314-232-9568).

Farrokh mentioned that all the papers from the Culham IAEA committee meeting in March would be refereed and published in a special Fusion Engineering and Design publication. He and Ian Cook will be guest editors. Copies of the papers should be provided at the meeting for the editors (three copies) and for the meeting participants. L. Waganer asked the status of the DOE nominations to the conference - Bill Dove is attempting to determine the quota of the number of people who may attend.

Farrokh also discussed his thoughts on the next project to be investigated by the ARIES group. He said there is high interest in investigating a neutron source and also looking at a (low-aspect ratio?) stellarator (application to be defined). The confinement concepts for the neutron source are completely open to all (magnetic) candidate concepts, including mirrors. Les Waganer asked about the hydrogen production application and Farrokh noted that John Sheffield, of ORNL, is starting a low-level project to evaluate the hydrogen production application. Farrokh would like to hear thoughts about the next ARIES project before the end of this week so he can prepare a consistent presentation

Plasma Physics

Parametric Trade Studies - Ron Stambaugh discussed some of the key points he made in a memo distributed on 14 February 1998, which documented his spreadsheet analysis results. The results highlighted plasma geometry and operating regimes that might offer potential benefit to increase the overall plant Q, mainly at an aspect ratio of 1.6 and a plasma triangularity similar to that of ARIES-ST. He was able to reproduce prior ARIES strawman results with 700 MW of alpha power as a benchmark case.

He then examined several changes in the plasma conditions and power core geometry as to their influence on the performance of the machine, which was established as the plant Q. The plant Q of the benchmark case was 2.19. The spreadsheet analysis held the fusion power as constant.

The first modification was to decrease the helium concentration from 15% to 5% based upon a t_He* / t_E of 5 (reference Phil West and ITER physics). This increased the plant Q from 2.19 to 2.58.
The second modification was to reduce the inboard shield thickness from 30 cm to 20 cm as agreed at the January ARIES project meeting. This produced a plant Q of 3.07.
The next change was to eliminate the inboard 10-cm scrape off-layer (SOL) and transfer that 10 cm to the centerpost, which raised the field on axis from 3.83 to 4.99 T. This change produced a plant Q of 3.64.
The final change was to increase the toroidal beta from 31% to 49%. This increased the plant Q to 4.69.

Ron Miller is reviewing Ron Stambaugh's results and trying to incorporate some of them in conjunction with the existing engineering and physics constraints. The present ARIES code results are based on a straight centerpost, which results in high centerpost resistive power losses. He tried to suppress the edge pedestal effects, but that change did not affect the results much.

Plasma Stability (GA) - Ron Stambaugh noted that GA is working on stability solutions for the aspect ratios of 1.4 and 1.6, elongation of 3.4, and triangularity of 0.6 which are marginally stable to ballooning modes. For the mode N = 2, the stabilizing wall should be at a radius of 1.3 times the plasma radius, while the N = 2 mode is stable with a wall at a radius of 1.1 to 1.2. With these parameters, Ron calculated a beta of 71% and a beta_N of 9.1%. There is no finite pressure gradient at the plasma edge. These results are similar to those in a GA LAR paper published last year. It was recommended that PPPL and GA exchange data sets and attempt to validate those results with each others codes. It was suggested that Ron Stambaugh should provide the code results in a EQ DISK file for Ron Miller to input into the ARIES Systems Code. But it was thought the only data needed would be the related PF coil data, which should be similar to the current data from PPPL. Thus, no new PF coil data should be necessary and Ron Miller and Ron Stambaugh will coordinate data to allow these results to be evaluated on the ARIES Systems Code. Ron Stambaugh suggested the triangularity be set at 0.6 for kink stability.

Plasma Stability (PPPL) - Chuck Kessel reported that they continue to struggle to define plasma regimes that are vertically stable. The long run times of the TSC code necessitated that Chuck rewrite portions of the code to improve the run times. Now he can more efficiently run cases to assess vertical stability conditions with varying conducting wall parameters and positions, both inboard and outboard. At present, Chuck is assuming conducting tungsten walls outboard behind the blanket, above and below the centerline, and at a distance of 0.5 times the plasma radius behind the first wall. These conditions generally result in a response time of 20-30 ms. Laila El-Guebaly has recommended placement of the conducting wall at a larger distance of 0.75 times the plasma radius. While Chuck has not assessed the 0.75 case, he is doubtful this would be a viable alternative. The group also discussed the possibility of placing a conducting wall in the inboard shield region. The steel wall would be ineffective due to the conductivity of the material. Tungsten has sufficient conductivity at elevated temperatures, but the proximity to the plasma environment significantly increases the afterheat and activation problems with tungsten. Alternate materials were proposed for more attractive properties.

Chuck is also using the PEST code, which assesses vertical plasma stability with ideal conducting shells, to help bound solutions assessed with the TSC code.

Heating and Current Drive - TK Mau provided preliminary engineering data on the low frequency, fast wave (LFFW) current drive antenna configuration. It uses 16 steel straps located on the outboard reactor midplane. These straps are contained in an enclosure on the first wall, which is approximately 35 cm radially thick. The total amount of steel is 0.05 m³. Laila El-Guebaly is assessing the impact of this amount of steel on the tritium breeding ratio.

Systems Studies

Ron Miller has posted, for evaluation, seven pre-strawman cases (A through F) on his web site . He has received feedback from the early sets (A through D) which helped him formulate some of the later cases. Plasma edge density pedestal profile has been included in some runs and not in others to determine its influence. RAMI modeling was included in Run E. The shield masses remain to be corrected. PF coil costs, which were missing on prior cases, now are included. The centerpost is now modeled as a straight element due to maintenance requirements. The baseline configuration of the centerpost shield is ferritic steel structure cooled with helium, and the pumping power for the both inboard shield and outboard blanket helium coolant has been included (approximately 120 MW). Run F shows the best results to date for a spherical tokamak power plant design.

Ron Miller noted that this run does not approach the plant Q of 4 which Ron Stambaugh had estimated. He reviewed the major reasons why the cost of electricity remains high for the spherical tokamak.

The straight centerpost, even with a 70% conductor filling fraction, has high resistive power losses. [Note: The 85% conductor filling fraction strawman had not been run to date.]
The helium pumping power losses associated with the centerpost cooling are very high; however, useful power is obtained from the cooling medium.
The thinning of the centerpost shield potentially could have helped increase the fusion power, instead that space was required for helium flow channels and a plenum.
A neon impurity was added to account for the necessary radiation to distribute the heat to the divertor and first wall. It was suggested that krypton would be a better impurity.
The inclusion of the PF coil masses and costs increased the capital cost and COE values.

The posted cases by Ron Miller had used the helium coolant for the inboard shield. Dai-Kai noted that both water and helium are possible coolants for the inboard shield. He has previously mentioned several reasons why lithium-lead would not be a suitable coolant for the inboard shield. Dai-Kai offered to do a trade study of these two coolants (pressure, pumping power, and shield thickness) providing the water case is acceptable from a neutronic standpoint.

Engineering

General - Les Waganer reported the general status of the engineering activities forwarded from Mark Tillack, who could not attend the conference call.

ST Workshop: Mark, and most others, were unaware of the ST workshop until yesterday, and will not have an opportunity to participate. This would also exclude helping others to prepare in case they want to participate. (During the meeting, Chuck Kessel said Steve Jardin may attend.)
IB Shield: Mark's opinion on the IB shield "controversy" is to retain He-cooled steel as our reference design approach unless and until we have an acceptable alternative. The examination of design options is at the heart of our study, and so any analysis of the relative merits and problems of water or LiPb should be adequately documented. Dai Kai and Laila should be, and are, working together on this problem.
Divertor Design: Regarding the divertor, Mark said he was sorry to say he had not made much progress. Mike Ulrickson and Clement Wong have been talking about studying a He-cooled tungsten divertor in the APEX project, and so, Mark is convinced that the APEX team is moving in a reasonable direction. Mike feels a credible design with high heat flux capability can be found. Mark's main contribution will be to examine the "extended surface" option, whereas Mark believes Mike favors the porous metal concept.
Divertor Configuration: At this point, there is no evidence that the FW can handle 1 MW/m² of surface heat flux; thus, Mark would be inclined to set the limit (optimistically) at 0.8 MW/m². If an optimistic peaking factor of 1.0 is assumed, then the split of power between the FW and divertor can be derived. The alpha power would be around 640-720 MW, depending on which strawman is preferred. The major radius would go up slightly, so that the FW area would be at least 400 m². If the 400 m² area is multiplied by 0.8 MW/m², a power level of 320 MW would be applied to the FW. For the lower value of alpha power, that's very close to a 50/50 split. Mark would feel much more comfortable if we can handle more than half the power in the divertor. If the 2 MW/m² value of peak divertor heat flux from ARIES-RS is adopted, then >160 m² of heated divertor surface area is needed. This requires an outboard slot length of heated surface between 1.5 to 2 m, assuming knowledge of how to distribute the power. If we have the ROM for a 2-m divertor slot, then it would help provide performance margin for the first wall.
APPLE: Mark suggested that Dai Kai summarize the status of the moving bed blanket concept, APPLE. Last week there was a half-day meeting on this subject. It appears to be a contender in the restructured advanced technology program. This concept was proposed as part of an ARIES study of advanced blanket concepts. At first, many people scoffed at it, but it is becoming more accepted in light of its merits relative to liquid surfaces (which have a substantial following now).

Blanket - Dai-Kai Sze noted the general consensus that the maximum heat flux capability for the outboard, helium-cooled first wall is 0.8 MW/m² (peak) for a structural temperature of 550°C. If the maximum structural temperature can be raised to 600°C, this would enhance the heat transfer capability and increase the heat flux capability.

In the inboard region, there is approximately 400 MW of thermal heat to remove. If the coolant is helium, at a reasonable pressure, the flow velocity is 200 m/s in the 20 cm annulus. If the pressure can be increased, the velocity and/or flow area can be decreased. If the centerpost is water-cooled, the heat transfer is much improved; but neutrons are absorbed in the water, which lessens the tritium breeding ratio. Beryllium can be added behind the first wall as a reflector, but this adds other concerns. Lithium-lead coolant has severe MHD problems in this area.

Neutronics and Shielding - Laila El-Guebaly reported that she distributed the radial build for the strawman to the team. She also has documented inboard radial builds for three inboard shield options. These are (A) FS/He (baseline), (B) FS/H₂O, and (C) LiPb/FS. A 10 cm space for helium manfold is needed to keep the pumping power within reasonable limits. This option runs hot and the heat is recovered. The water coolant option offers the thinnest shield option; but running reasonably cool, no heat is recovered and the outboard breeding will drop to an unacceptable level. Lithium lead has the most technical problems. She also noted that the posted strawmen A and B have the same COE meaning that reducing the inboard shield thickness by 10 cm does not have major impact on the overall cost.

Safety - Hesham Khater reported that he has been working on the helium-cooled inboard shield. The waste disposal rating for this option is acceptable.

Maintenance and Configuration - Xueren Wang said that he has been working on sizing the hydraulic cylinders and equipment to raise and lower the power core. He is examining the design approach to coordinate the several pistons to uniformly lift the entire power core. He is also sizing the removable support structure used in the operational phase of the plant.