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- Accelerator-driven systems; Monte Carlo N-Particle eXtended (MCNPX); Nuclear reactions – Computer simulation; Radioactive wastes — Transmutation; Spallation (Nuclear physics) (1)
- Laser beams; Monte Carlo N-Particle eXtended (MCNPX); Nuclear reactions – Computer simulation; Radioactive wastes — Transmutation; Spallation (Nuclear physics) (1)
- Monte Carlo N-Particle eXtended (MCNPX); Neutron flux; Neutrons; Nuclear reactors; Particles (Nuclear physics); Radioactive wastes — Transmutation; Spallation (Nuclear physics); Spent reactor fuels; Transmutation (Chemistry) (1)
Articles 1 - 3 of 3
Full-Text Articles in Physical Sciences and Mathematics
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth, Denis Beller
Reactor Campaign (TRP)
The second year of this project involved modeling several aspects of the LANCSE beam experiments:
- Modeling targets of varying diameter in air, in a vacuum, and in the presence of humid air;
- Modeling various proton beam profiles;
- Modeling the effects of off-axis proton beam impingement on the target;
- Modeling the asymmetry introduced by the steel table below the target;
- Modeling the effect of varying ratios of Pb to Bi and the effect of impurities; and
- Modeling the system, including other structures within the test room.
With the experience gained through modeling these systems, the UNLV researchers plan, with the assistance …
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth
Radiation Transport Modeling Using Parallel Computational Techniques, William Culbreth
Reactor Campaign (TRP)
The Advanced Fuel Cycle Initiative (AFCI) program will rely on the use of accurate calculations and simulations of criticality and shielding for the separation process of the longlived isotopes that present a significant safety hazard in commercial spent fuel. To help design and verify the safety of the separation process, the neutronics code MCNPX will be used to model the distribution of neutron flux within the fuel blanket and to determine the neutron multiplication, keff. However, the cross section libraries and computational methods used by MCNPX at these neutron energies still have some uncertainty and will require validation. …
Monte Carlo Verification And Modeling Of Lead-Bismuth Spallation Targets, Daniel R. Lowe
Monte Carlo Verification And Modeling Of Lead-Bismuth Spallation Targets, Daniel R. Lowe
Reactor Campaign (TRP)
No abstract provided.