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MCODE, Version 2.2: An MCNP-ORIGEN DEpletion Program
(Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2008-12)
MCODE Version 2.2 is a linkage program, which combines the continuous-energy
Monte Carlo code, MCNP-4C, and the one-group depletion code, ORIGEN2, to perform
burnup calculations for nuclear fission reactor systems. MCNP ...
Flexible Conversion Ratio Fast Reactor Systems Evaluation Final Report
(Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2008-06)
Executive Summary:
The goal of this project is to develop the conceptual designs of fast flexible conversion
ratio reactors using lead and liquid salt coolants and to compare the results with a gascooled fast reactor ...
A System Dynamics Study of the Nuclear Fuel Cycle with Recycling: Options and Outcomes for the US and Brazil
(Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2008-11)
A system dynamics simulation technique is applied to generate a new version of the
CAFCA code to study mass flows in the nuclear fuel cycle, and the impact of different
options for advanced reactors and fuel recycling ...
Thermal Hydraulic Analysis of a Low Enrichment Uranium Core for the MIT Research Reactor
(Massachusetts Institute of Technology. Center for Advanced Nuclear Energy Systems. Nuclear Fuel Cycle Program, 2008-01)
The MIT research reactor (MITR) is converting from the existing high enrichment
uranium (HEU) core to a low enrichment uranium (LEU) core using a high-density
monolithic UMo fuel. The design of an optimum LEU core for ...