An Assessment Of Critical Thermal-Hydraulic Problems in A Deuterium-Tritium Solid Breeder Blanket-Reference-Cited by-同舟云学术

An Assessment Of Critical Thermal-Hydraulic Problems in A Deuterium-Tritium Solid Breeder Blanket

Published:1983-09 Issue:2P1 Volume:4 Page:253-262
ISSN:0272-3921
Container-title:Nuclear Technology - Fusion
language:en
Short-container-title:Nuclear Technology - Fusion

Author:

Misra Balabhadra¹,Clemmer Robert G.¹,Smith Dale L.¹

Affiliation:

1. Argonne National Laboratory, Fusion Power Program, 9700 South Cass Avenue, Argonne, Illinois 60439

Publisher

Informa UK Limited

Subject

General Engineering

Link

https://www.tandfonline.com/doi/pdf/10.13182/FST83-A22817

Reference13 articles.

1. J. R. POWELL et al. “Studies of Fusion Reactor Blankets with Minimum Radioactive Inventory and Tritium Breeding in Solid Lithium Compounds, A Preliminary Report,” BNL-18236, Brookhaven National Laboratory (1973).

2. K. SAKO et al. “Conceptual Design of a Gas-Cooled Tokamak Reactor,” JAERI-M, Japan Atomic Energy Research Institute (1973).

3. K. OKULA and D. K. SZE, “Tritium Recovery from Solid Breeders: Implications of Existing Data,” UWFDM-351, University of Wisconsin (1980).

4. R. WISWALL and E. WIRSING, “The Removal of Tritium from Fusion Reactor Blankets,” BNL-50748, Brookhaven National Laboratory (1977).

5. D. W. KEARNY et al. “Conceptual Design Study of a Non-Circular Tokamak Demonstration Fusion Power Reactor,” GA-A13992, General Atomic Company (1976).

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