|(UCSD)||Miller, Najmabadi, Tillack, Wang|
There was a brief discussion of the people planning to attend the IAEA Technical Committee Meeting on Fusion Power Plant Design to be held in Culham 23-27 March 1998. The 13th ANS Topical Meeting on the Technology of Fusion Energy is scheduled for June 1998 with summaries due on 9 January. Dai Kai Sze noted that an engineering workshop might be held the week following the ANS meeting that would focus on advanced blankets, tritium systems, and plasma materials interaction.
It was suggested the dates for the next ARIES meeting be extended to start the meeting at 1:00 on 7 January (Wednesday) and continue through noon Friday, 9 January. L. Waganer canvassed the ARIES team the following day and all respondents agreed with the new dates, which were adopted. L. Waganer will prepare an agenda for the meeting. The next conference call is scheduled for 17 December 1997.
A question was raised that the Physics group should affirm the assumption that 20% of the radiative power should be intercepted by the inboard divertor (after the 50-50 split of the radiative power between the divertor and the wall.)
Dai Kai Sze explained that the blanket selection is dependent on three issues:
All first wall and blanket systems to be recommended must satisfy all the design requirements for the blanket system. All known blanket systems have some issues to be addressed and quantified, both in terms of the blanket alone as well as the other interacting systems and the plant as a whole. For the present, we are planning to concentrate on the Li-Pb cooled ferritic steel blanket while retaining the Li/V blanket as the reference case. Since the blanket lifetime would have impact on the overall economics, Dai-Kai suggested enabling the availability routine in the systems code to evaluate the blanket on that parameter. March is the timeframe when the choice of the blanket should be accomplished.
Centerpost - Wayne Reiersen could not participate in the call due to the NSTX meeting.
Configuration and Maintenance - The power core configuration is being modeled in the CATIA format. The divertor is being modeled at the present time. Vertical maintenance seems to be the best option. Don Steiner noted that Wayne Reiersen is analyzing the demountable TF coil joints that are at a radius sufficient for the downward vertical maintenance.
Safety Studies (LOCA) - H. Khater reported that E. Mogahed has completed a LOCA analysis involving the power core decay heat. He is still waiting for confirmation regarding mechanical approach to enhance the conduction heat transfer from the inboard shield to the copper centerpost (Action: W. Reiersen). He modified the code to add a conduction term equivalent to 1% steel conductivity.
Nuclear Parameters and Radial Build - Laila El-Guebaly stated that she is working on the neutronics of the inboard and outboard assuming a homogeneous LiPb-cooled blanket. She is also working on the technique to convert from a 1-D to a 3-D neutronics analysis.
Divertor - Mark Tillack reviewed the basic power distribution assumptions: 50% of the total radiation power to the divertor, of which 20-25% of the divertor power will be directed to the inboard divertor surfaces (see Physics question.) With the present plasma configuration, this amounts to a surface heat flux on the inboard divertor surfaces similar to the ARIES-RS divertor, which is reasonable. We will likely assume a radiative divertor. The outboard slot is around 1.5 m long using tungsten. This is plenty of surface area, but the inboard region and local shielding for the centerpost are problem areas. Mark is favoring the use of a straight centerpost to get sufficient area for the inboard divertor slot. If the centerpost is straight, it may have to be shorter to help bring the joule losses within reason.