A Computational Method for Neutron Transport Problems in Toroidal Geometry
Author:
Affiliation:
1. Argonne National Laboratory, Applied Physics Division, 9700 South Cass Avenue, Argonne, Illinois 60439
Publisher
Informa UK Limited
Subject
Nuclear Energy and Engineering
Link
https://www.tandfonline.com/doi/pdf/10.13182/NSE78-A27131
Cited by 6 articles. 订阅此论文施引文献 订阅此论文施引文献，注册后可以免费订阅5篇论文的施引文献，订阅后可以查看论文全部施引文献
1. DEVELOPMENT OF RADIATION TRANSPORT CODE IN 3-DIMENSIONAL (X, Y, Z) GEOMETRY FOR SHIELDING ANALYSES BY DIRECT INTEGRATION METHOD;J NUCL SCI TECHNOL;1987
2. Development of Radiation Transport Code in Three-Dimensional (X, Y, Z)Geometry for Shielding Analyses by Direct Integration Method;Journal of Nuclear Science and Technology;1987-03
3. Neutronic Data-Base Assessment for U.S. INTOR;Fusion Technology;1986-11
4. Nuclear Design and Analysis of A Deuterium-Deuterium Tokamak Reactor, Wildcat;Nuclear Technology - Fusion;1983-07
5. Source-to-Incident Flux Relation for a Tokamak Fusion Test Reactor Blanket Module;Nuclear Technology - Fusion;1982-07
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