Shielding analysis for the ITER divertor and vacuum pumping ducts
- Univ. of Wisconsin, Madison, WI (United States)
- Max-Planch-Institut fur Plasmaphysik, Garching bei Munchen (Germany)
In the Engineering Design Activity (EDA) of the International Thermonuclear Experimental Reactor (ITER), reducing nuclear heating in the toroidal field (TF) coils to acceptable levels particularly in the regions behind the divertor and adjacent to the divertor vacuum pumping ducts has been identified as an important shielding issue. The poloidal distribution of the neutron wall loading in the different regions of ITER has been determined using the Monte Carlo code MCNP. The detailed geometrical configuration of the ITER first wall has been modeled in the calculation. Source neutrons are sampled from the plasma zone according to the fusion power density distribution within the single null plasma shape. The average neutron wall loading is 0.913 MW/m{sup 2} for the nominal 1500 MW fusion power. The calculated average neutron wall loadings at the outboard and inboard first walls, are 1.044 and 0.735 MW/m{sup 2}, respectively. The peak outboard and inboard neutron wall loadings are 1.193 and 0.923 MW/m{sup 2}, respectively. The peak neutron wall loading in the divertor region is 0.559 MW/m{sup 2} at the upper surface of the middle divertor plate facing the plasma x-point.
- OSTI ID:
- 197078
- Report Number(s):
- CONF-940664-; TRN: 95:005767-0256
- Resource Relation:
- Conference: ISFNT-3: international symposium on fusion nuclear technology, Los Angeles, CA (United States), 27 Jun - 1 Jul 1994; Other Information: PBD: 1994; Related Information: Is Part Of Third international symposium on fusion nuclear technology; PB: 362 p.
- Country of Publication:
- United States
- Language:
- English
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