A5059809211- Nuclear reactor physics and engineering
- Nuclear Physics and Applications
- Nuclear and radioactivity studies
- Manufacturing Process and Optimization
- Gas Dynamics and Kinetic Theory
- Matrix Theory and Algorithms
- Graphite, nuclear technology, radiation studies
- Nuclear Materials and Properties
- Medical Imaging Techniques and Applications
- Probabilistic and Robust Engineering Design
- Image and Signal Denoising Methods
- Radiation Therapy and Dosimetry
- Numerical methods in inverse problems
- Advanced Numerical Methods in Computational Mathematics
- Sparse and Compressive Sensing Techniques
- Quantum, superfluid, helium dynamics
- Atomic and Subatomic Physics Research
- Radiative Heat Transfer Studies
- Heat Transfer and Optimization
- Advanced X-ray and CT Imaging
- Computational Geometry and Mesh Generation
- Granular flow and fluidized beds
- Superconducting Materials and Applications
- Innovations in Concrete and Construction Materials
Los Alamos National Laboratory
2012-2025
Texas A&M University
2014-2016
After several years of effort involved in merging the Los Alamos National Laboratory MCNP5 and MCNPX codes, 2013 first production release version 6 Monte Carlo N-Particle ® , or MCNP code MCNP6.1 was distributed publicly. Since then, three significant releases have been issued: MCNP6.1.1beta 2014, MCNP6.2 2018, MCNP6.3 2023. While each always contains new features, enhancements, bug fixes, has had a different primary focus, ranging from improved calculational efficiency to powerful utilities...
In this work, we present a methodology for using compressed sensing ideas to reduce the memory footprint of Monte Carlo simulation where scalar flux over entire problem is desired. We first introduce idea disjoint tallies, i.e., particle tallies that are not contiguous in space. Then give prescription choosing these randomly and performing local reconstruction small number based on minimizing total variation (TV) norm reconstruction. Results TRIGA reactor indicate our method can accurate...
This presentation details some of the new features MCNP6.3. With an included look at Deimos Project. Key points include unstructured mesh and HDF5, fission matrix for criticality calculations, tally features. Future work a Q&A are provided end.
This paper describes how to directly sample Monte Carlo flight paths simulate particle transport in media with continuously varying cross sections without using iterative methods. Specifically, this approach supports linear finite-element codes tetrahedral meshes. Two analytical test problems are used verify the method.
It is well known that in an infinite, source-free, purely scattering medium with physically realizable cross sections, neutrons attain a Maxwell-Boltzmann distribution characterized by the material temperature. In this work we look at how small variations to these conditions change behavior of thermal-neutron space and energy. Specifically, our analysis examines influence amounts absorption source as temperature material. We restrict study regions away from boundary initial layers. The...