Michael Rising
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View article: ENDF/B-VIII.1: Updated Nuclear Reaction Data Library for Science and Applications
ENDF/B-VIII.1: Updated Nuclear Reaction Data Library for Science and Applications Open
The ENDF/B-VIII.1 library is the newest recommended evaluated nuclear data file by the Cross Section Evaluation Working Group (CSEWG) for use in nuclear science and technology applications, and incorporates advances made in the six years s…
View article: Functional Expansion Tallies Using MCNP6 PTRAC Files [Slides]
Functional Expansion Tallies Using MCNP6 PTRAC Files [Slides] Open
View article: Nuclear data needs associated with criticality monitoring during Fukushima Daiichi fuel debris retrieval operations [Poster]
Nuclear data needs associated with criticality monitoring during Fukushima Daiichi fuel debris retrieval operations [Poster] Open
View article: FSEN Reaction Rate Calculations in MCNP [Slides]
FSEN Reaction Rate Calculations in MCNP [Slides] Open
View article: MCNP6 Developments: A 2024-25 Year in Review [Slides]
MCNP6 Developments: A 2024-25 Year in Review [Slides] Open
View article: Challenges in Explicit Modeling of TRISO in MCNP [Slides]
Challenges in Explicit Modeling of TRISO in MCNP [Slides] Open
View article: 1FRAME: 1F Fuel Retrieval and Monitoring Experiments [Slides]
1FRAME: 1F Fuel Retrieval and Monitoring Experiments [Slides] Open
View article: Neutron Leakage Spectra Sensitivity Simulations for PARADIGM [Poster]
Neutron Leakage Spectra Sensitivity Simulations for PARADIGM [Poster] Open
View article: Direct Sampling of Monte Carlo Flight Paths in Tetrahedral Meshes with Linear Finite-Element Cross Sections
Direct Sampling of Monte Carlo Flight Paths in Tetrahedral Meshes with Linear Finite-Element Cross Sections Open
View article: The MCNP<sup>®</sup>6 code: A decade of progress
The MCNP<sup>®</sup>6 code: A decade of progress Open
After several years of effort involved in merging the Los Alamos National Laboratory MCNP5 and MCNPX codes, in 2013 the first production release of version 6 of the Monte Carlo N-Particle ® , or MCNP ® , code MCNP6.1 was distributed public…
View article: Derivation and verification of the direct-sampling method for simulating Monte Carlo flight paths in tetrahedral meshes with linear finite-element cross sections
Derivation and verification of the direct-sampling method for simulating Monte Carlo flight paths in tetrahedral meshes with linear finite-element cross sections Open
This paper provides a derivation of a direct-sampling approach for modeling continuously varying cross sections in tetrahedral-mesh-based Monte Carlo codes. Specifically, cross sections are spatially approximated using linear nodal finite …
View article: FY24 MCNP(R) Updates for the Nuclear Criticality Safety Program [Slides]
FY24 MCNP(R) Updates for the Nuclear Criticality Safety Program [Slides] Open
not provided.
View article: Whisper updates [Slides]
Whisper updates [Slides] Open
not provided.
View article: Cu covariance study including impacts on past (and future) Zeus experiments
Cu covariance study including impacts on past (and future) Zeus experiments Open
This presentation covers why we should care about Cu nuclear data. It also covers the current state of Cu63 and Cu65 covariances and the Impact on specific critical experiments.
View article: The MCNP®6 Code: A Decade of Progress
The MCNP®6 Code: A Decade of Progress Open
not provided.
View article: Visualizing Whisper computation of the calculational margin [Slides]
Visualizing Whisper computation of the calculational margin [Slides] Open
View article: MCNP6 Developments: A 2023-24 Year in Review [Slides]
MCNP6 Developments: A 2023-24 Year in Review [Slides] Open
View article: Analytic Sensitivity Coefficients for General Multigroup Infinite Medium k-Eigenvalue Problems
Analytic Sensitivity Coefficients for General Multigroup Infinite Medium k-Eigenvalue Problems Open
View article: LANL-SNL Collaboration on NCS Validation.
LANL-SNL Collaboration on NCS Validation. Open
During 2016, nuclear criticality safety (NCS) practitioners from SNL and code developers from LANL collaborated in several areas of interest to the DOE/NNSA Nuclear Criticality Safety Program (NCSP). This collaboration involved. Testing of…
View article: Studying the Random Number Generators in MCNP6 using an Analytic Benchmark
Studying the Random Number Generators in MCNP6 using an Analytic Benchmark Open
An analytic solution to a previously studied toy problem is derived and used as a code verification benchmark. Using various Random Number Generators (RNGs) in MCNP6, including the newest SFC64 RNG available in MCNP6.3.1, and their various…
View article: The MUSIC Critical Benchmark and Nuclear Data
The MUSIC Critical Benchmark and Nuclear Data Open
The Measurement of Uranium Subcritical and Critical (MUSIC) experiment was a series of measurements of critical and subcritical configurations of bare highly enriched uranium. The goal was to compare measurement methods, analysis technique…
View article: Fixed Source Sensitivity Calculations for Inertial Confinement Fusion Applications
Fixed Source Sensitivity Calculations for Inertial Confinement Fusion Applications Open
A numerical code library was developed for the radiation transport code MCNP6.3 to calculate generalized response sensitivity coefficients for fixed source neutron transport problems with applications to inertial confinement fusion (ICF) e…
View article: MCNP6(R) New Features and Improvements for Reactor Physics Applications
MCNP6(R) New Features and Improvements for Reactor Physics Applications Open
View article: A Delta-Tracking Method That Supports Finite-Element Material Properties in Mesh-Based Monte Carlo Codes
A Delta-Tracking Method That Supports Finite-Element Material Properties in Mesh-Based Monte Carlo Codes Open
View article: Implementation and Verification of Element-Wise Density and Temperature Specifications in MCNP6 Unstructured Mesh Simulations
Implementation and Verification of Element-Wise Density and Temperature Specifications in MCNP6 Unstructured Mesh Simulations Open
View article: Nuclear Data Sensitivity/Uncertainty Studies of Tantalum-Reflected Systems
Nuclear Data Sensitivity/Uncertainty Studies of Tantalum-Reflected Systems Open
Tantalum (Ta) is a refractory metal with high melting point and high corrosion resistance. This makes it useful as a mold material for plutonium casting operations. Unfortunately, there are few criticality benchmarks with high nuclear data…
View article: MCNP<sup>®</sup> Code Version 6.3.1 Theory & User Manual
MCNP<sup>®</sup> Code Version 6.3.1 Theory & User Manual Open
This document acts as a repository of knowledge for the Monte Carlo N-Particle (MCNP) transport computer code. It is maintained alongside the source code and attempts to introduce new users and re-familiarize experienced users with the the…
View article: MCNP6® New Features and Improvements for Reactor Physics Applications [Slides]
MCNP6® New Features and Improvements for Reactor Physics Applications [Slides] Open
This presentation details some of the new features of MCNP6.3. With an included look at the Deimos Project. Key points include unstructured mesh and HDF5, fission matrix for criticality calculations, and new tally features. Future work and…
View article: MCNP6® Recent Updates on Modernization, R&D, and Release Plans
MCNP6® Recent Updates on Modernization, R&D, and Release Plans Open
View article: Sensitivity Tool Needs for Modern Nuclear Data Validation
Sensitivity Tool Needs for Modern Nuclear Data Validation Open