Time dependent burn-up and fission products inventory calculations in the discharged fuel of the Syrian MNSR

Author:

Omar H.,Ghazi N.

Publisher

Elsevier BV

Subject

Nuclear Energy and Engineering

Reference14 articles.

1. J.F. Briesmeister (Ed.), 2000. MCNP – A general Monte Carlo N-Particle Transport Code, Version 4C. LA-13709-M, Los Alamos National Laboratory.

2. Croff, A.G., 1980. A user’s Manual for the ORIGEN 2 computer code. ORNL/TM7175 Oak Ridge National Laboratory, July 1980.

3. Burn up calculations for the Iranian miniature reactor: a reliable and safe research reactor;Faghihi;Nuclear Engineering and Design,2009

4. Fowler, T.B., Vondy, D.R., 1971. Nuclear reactor core analysis code CITATION. ORNL-TM-2496.

5. Core conversion analyses of the Syrian MNSR reactor from HEU to LEU and MEU fuel with homogeneously mixed burnable poisons;Ghazi;Applied Radiation and Isotopes,2009

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