Determination of the equilibrium composition of cores with continuous fuel feed and removal using MOCUP

Joint International Topical Meeting on Mathematics and Computations and Supercomputing in Nuclear Applications, M and C + SNA 2007

Volumn , Issue , 2007, Pages

  Fratoni, Massimiliano ^a   Greenspan, Ehud ^a  

^a UNIVERSITY OF CALIFORNIA   (United States)

Author keywords

Depletion; MCNP; MOCUP; Molten salt; ORIGEN; Pebble

Indexed keywords

DEPLETION; FUEL FEEDING; MOCUP; MOLTEN SALT REACTORS; PEBBLE BED REACTORS; RESIDENCE TIME;

ALGORITHMS; CORE MELTDOWN; NUCLEAR FUELS;

REACTOR CORES;

EID: 84858463552     PISSN: None     EISSN: None     Source Type: Conference Proceeding    
DOI: None     Document Type: Conference Paper

References (9)

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2. X-5 Monte Carlo Team, "MCNP - A General Monte Carlo N-Particle Transport Code, Version 5," LANL 2003
3. Croff, A.G., "A User Manual for the ORIGEN2 Computer Code," ORNL/TM-7175
4. Croff, A.G., "ORIGEN2: A Versatile Computer Code for Calculating the Nuclide Compositions and Characteristics of Nuclear Materials," Nuclear Technology, Vol. 62, September 1983
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6. Gauld, I.C., Murphy, B.D. and Williams M. L., "ORIGEN-S Data Libraries," ORNL/TM-2005/39 Version 5 Vol. III, Sect. M6, April 2005
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8. Forsberg, C. W., Pickard, P. and Peterson RF., "Molten-Salt-Cooled Advanced High-Temperature Reactor for Production of Hydrogen and Electricity," Nuclear Technology, Vol. 144, pp. 289-302, 2003
9. de Zwaan, S.J., "The Liquid Salt Pebble Bed Reactor," PNR-131-2005-008, Delft University of Technology, The Netherlands, November 2005