TY - GEN
T1 - Scoping studies for small steady-state tokamaks for divertor testing
AU - Galambos, J. D.
AU - Peng, Y. K.M.
AU - Nelson, B. E.
AU - Hirshman, S. P.
AU - Fogarty, P. J.
N1 - Publisher Copyright:
© 1991 IEEE.
PY - 1991
Y1 - 1991
N2 - A prime uncertainty in next-generation devices is the divertor performance. For the International Thermonuclear Experimental Reactor (ITER), the divertor limit often plays a more critical role in the operational scenario definition than do beta limit and energy confinement constraints. Hence, it is desirable to test the divertors in an environment as close as possible to that expected in next-step burning plasma experiments. Initial global scoping studies are done for small, steady-state, copper coil, beam-driven tokamaks that are dedicated to divertor testing. The usual ITER global physics models (beta limit, energy confinement, and analytic divertor heat load calculation) are incorporated, and for performance criteria we require that the divertor heat load and plasma collisionality in the edge region be similar to those expected in ITER. The smallest, lowest-cost devices satisfying these constraints tend to have major radius below 1 m, plasma current of 0.5 to 1 MA, and low aspect ratio and costs of a few tens of millions of dollars. Injection powers of about 4 to 5 MW are needed to sustain the plasma current, maintain plasma power balance, and provide the required divertor heat load.
AB - A prime uncertainty in next-generation devices is the divertor performance. For the International Thermonuclear Experimental Reactor (ITER), the divertor limit often plays a more critical role in the operational scenario definition than do beta limit and energy confinement constraints. Hence, it is desirable to test the divertors in an environment as close as possible to that expected in next-step burning plasma experiments. Initial global scoping studies are done for small, steady-state, copper coil, beam-driven tokamaks that are dedicated to divertor testing. The usual ITER global physics models (beta limit, energy confinement, and analytic divertor heat load calculation) are incorporated, and for performance criteria we require that the divertor heat load and plasma collisionality in the edge region be similar to those expected in ITER. The smallest, lowest-cost devices satisfying these constraints tend to have major radius below 1 m, plasma current of 0.5 to 1 MA, and low aspect ratio and costs of a few tens of millions of dollars. Injection powers of about 4 to 5 MW are needed to sustain the plasma current, maintain plasma power balance, and provide the required divertor heat load.
UR - http://www.scopus.com/inward/record.url?scp=85067662824&partnerID=8YFLogxK
U2 - 10.1109/FUSION.1991.218676
DO - 10.1109/FUSION.1991.218676
M3 - Conference contribution
AN - SCOPUS:85067662824
T3 - Proceedings - Symposium on Fusion Engineering
SP - 1114
EP - 1118
BT - Proceedings - 14th IEEE/NPSS Symposium Fusion Engineering, FUSION 1991
PB - Institute of Electrical and Electronics Engineers Inc.
T2 - 14th IEEE/NPSS Symposium Fusion Engineering, FUSION 1991
Y2 - 30 September 1991 through 3 October 1991
ER -