The nuclear fuel cycle: analysis and management
American Nuclear Society, 1990 - Mathematics - 378 pages
This presents a balanced overview of the major methods currently available for obtaining numerical solutions in neutron and gamma-ray transport. It focuses on methods particularly applicable to the complex problems encountered in the analysis of reactors, fusion devices, radiation shielding, and other nuclear systems. It will be valuable as a self-contained reference and text to practicing engineers involved in research and development, to users of large transport computer codes for engineering analysis, and to first-year graduate students.
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Nuclear Fuel Resources
Enrichment and Conversion
Reactor Fuel Design and Fabrication
9 other sections not shown
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absorption activities amount Atomic average batch boron burnable poisons burnup calculation capacity factor cask chemical cladding concentration contains control rod coolant core cost cross section decay decommissioning depletion depreciation developed disposal effects electricity energy enrichment equation example facility fissile fission products fraction fresh fuel fuel assemblies fuel cycle Fuel Management fuel rods gaseous diffusion geologic heat irradiated isotopes LWRs metal method MHTGR MOX fuel multigroup MWD/MTU NASAP natural uranium neutron flux Nucl nuclear fuel nuclear fuel cycle nuclear power plant Nuclear Reactor nuclides NWPA package pellets plutonium pressure vessel problem radiation radioactive materials radioactive wastes reactivity recycling reduce refueling repository reprocessed uranium reprocessing result shown in Fig shutdown spent fuel storage surface Table tails taxes Technol temperature thermal thorium Trans transportation United uranium and plutonium utility