Analytic Sensitivity Coefficients for General Multigroup Infinite Medium k-Eigenvalue Problems
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analytic, sensitivity coefficient, multigroup, inf↗
Engineering topics
Vaquer, Pablo Andres
Publications and source records attributed to Vaquer, Pablo Andres.
Using ParaView to Visualize MCNP6 Fission Matrix Eigenmodes
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MCNP, ParaView, Python, fission, matrix, eigenmode↗
MCNP6(R) New Features and Improvements for Reactor Physics Applications
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Rising, Michael Evan↗
Implementation and Verification of Element-Wise Density and Temperature Specifications in MCNP6 Unstructured Mesh Simulations
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Monte Carlo, unstructured mesh, elementwise, tempe↗
A Delta-Tracking Method That Supports Finite-Element Material Properties in Mesh-Based Monte Carlo Codes
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Monte Carlo, delta tracking, finite element, mesh,↗
Implementation and Verification of Element-Wise Density and Temperature Specifications in MCNP6 Unstructured Mesh Simulations
Explore the source record for details and available documents.
Monte Carlo, continuously-varying, density, temper↗
A Delta-Tracking Method That Supports Finite-Element Material Properties in Mesh-Based Monte Carlo Codes
Explore the source record for details and available documents.
Monte Carlo, delta tracking, finite element, mesh,↗
Analytic Sensitivity Coefficients for General Multigroup Infinite Medium k-Eigenvalue Problems
This work presents a general set of equations that can be used to rapidly generate new benchmarks to verify nuclear data sensitivity calculations. The general multigroup infinite medium k-eigenvalue neutron transport equation is used to derive analytic expressions for the infinite medium k-eigenvalue, the scalar neutron flux and adjoint flux, and the sensitivity of k∞ to perturbations in the multigroup nuclear data of a single species, isotropic and elastic scattering, material. The multigroup nuclear data for U-235 and U-238 is presented along with their corresponding k-eigenvalues, forward flux, adjoint flux, and sensitivity profiles, which include the sensitivity of k∞ to the total, fission, capture, and scattering macroscopic cross sections as well as to the group-to-group scattering cross section matrix, group-wise fission neutron production, and the unconstrained and constrained fission neutron energy distribution.
73 NUCLEAR PHYSICS AND RADIATION PHYSICS↗