A5032192554- Nuclear reactor physics and engineering
- Nuclear Physics and Applications
- Graphite, nuclear technology, radiation studies
- Nuclear and radioactivity studies
- Nuclear Materials and Properties
- Radioactive contamination and transfer
- Radiation Detection and Scintillator Technologies
- Radioactive element chemistry and processing
- Medical Imaging Techniques and Applications
- Magnetic confinement fusion research
- Particle Detector Development and Performance
- Radiation Therapy and Dosimetry
- Spacecraft and Cryogenic Technologies
- Risk and Safety Analysis
- Fusion materials and technologies
- Advanced Data Storage Technologies
- Radiative Heat Transfer Studies
- Superconducting Materials and Applications
- Gas Dynamics and Kinetic Theory
- Radioactivity and Radon Measurements
- Laser-Plasma Interactions and Diagnostics
- Advanced Malware Detection Techniques
- Nuclear physics research studies
- Nuclear Engineering Thermal-Hydraulics
- Network Security and Intrusion Detection
Los Alamos National Laboratory
2008-2019
University of Florida
2005-2010
MCNP6 is simply and accurately described as the merger of MCNP5 MCNPX capabilities, but it much more than sum those two computer codes. result five years effort by code development teams. These groups people, residing in Los Alamos National Laboratory's (LANL) X Computational Physics Division, Monte Carlo Codes Group (XCP-3), Decision Applications Radiation Transport Team (D-5), respectively, have combined their efforts to produce next evolution MCNP. While maintenance bug fixes will...
MCNPX (Monte Carlo N‐Particle eXtended) is a general‐purpose Monte radiation transport code with three‐dimensional geometry and continuous‐energy of 34 particles light ions. It contains flexible source tally options, interactive graphics, support for both sequential multi‐processing computer platforms. based on MCNP4c has been upgraded to most MCNP5 capabilities. MCNP highly stable tracking neutrons, photons electrons, using evaluated nuclear data libraries low‐energy interaction...
(2010). The Enhancements and Testing for the MCNPX 2.6.0 Depletion Capability. Nuclear Technology: Vol. 170, Special Issue on 2008 International Congress Advances in Power Plants, pp. 68-79.
MCNP6 is simply and accurately described as the merger of MCNP5 MCNPX capabilities, but it much more than sum these two computer codes. result six years effort by code development teams. These groups people, residing in Los Alamos National Laboratory’s X Computational Physics Division, Monte Carlo Codes Group (XCP-3) Nuclear Engineering Nonproliferation Radiation Transport Modeling Team (NEN-5) respectively, have combined their efforts to produce next evolution MCNP. While maintenance major...
As advanced reactor concepts challenge the accuracy of current modeling technologies, a higher-fidelity depletion calculation is necessary to model time-dependent core reactivity properly for accurate cycle length and safety margin determinations. The recent integration CINDER90 into MCNPX Monte Carlo radiation transport code provides completely self-contained Carlo-linked capability. Two advances have been made in latest capability based on problems observed prereleased versions:...
We present neutronics calculations for a hypothetical fusion reactor based on the repetitively pulsed concept of plasma-jet-driven magneto-inertial (PJMIF). A PJMIF is envisioned to have replaceable, 3-m-radius spherical metal first wall exposed 14.1-MeV neutrons; fast-flowing FLiBe liquid blanket (with thickness 0.75 m) behind serving as primary coolant and tritium-breeding medium; finally an outer structural shielded by blanket. Cylindrical penetrations through both walls flowing allow...