Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design | Zendy

C.A. Wemple | Zendy; Bruce Schnitzler | Zendy; J.M. Ryskamp | Zendy

Open Access

Neutronics methods, models, and applications at the Idaho National Engineering Laboratory for the advanced neutron source reactor three-element core design

Author(s) -

C.A. Wemple,

Bruce Schnitzler,

J.M. Ryskamp

Publication year - 1995

Publication title -

osti oai (u.s. department of energy office of scientific and technical information)

Language(s) - English

Resource type - Reports

DOI - 10.2172/196552

Subject(s) - neutron transport , nuclear engineering , monte carlo method , neutron , core (optical fiber) , neutron source , nuclear physics , nuclear reactor core , nuclear data , computer science , physics , engineering , mathematics , telecommunications , statistics

A summary of the methods and models used to perform neutronics analyses on the Advanced Neutron Source reactor three-element core design is presented. The applications of the neutral particle Monte Carlo code MCNP are detailed, as well as the expansion of the static role of MCNP to analysis of fuel cycle depletion calculations. Results to date of these applications are presented also. A summary of the calculations not yet performed is also given to provide a {open_quotes}to-do{close_quotes} list if the project is resurrected

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