Publications
Showing 60 results for Author: Gregory G. Davidson
Jan, 2026
Journal
Automated Hybrid Variance Reduction on Advanced Architectures in the Shift Monte Carlo Code
Monte Carlo transport methods are the most accurate schemes for solving problems with complex energy and spatial features, but they come with a high computational cost. Although hybrid methods have enabled the use of Monte Carlo transport for a large class of problems, they still require significant computing resources. Modern multicore CPUs with large numbers of compute c…
Sep, 2025
Journal
The Transient Multi-Level method for Monte Carlo reactor statics calculations
The Transient Multi-Level (TML) method is applied to a time-dependent Monte Carlo transport solver to offload some of the computational burden of the expensive Monte Carlo solve to lower-order Coarse Mesh Finite Difference (CMFD) and Exact Point Kinetics Equations (EPKE) solvers via factorization of the neutron flux at the transport and CMFD levels using the Predictor Corr…
Feb, 2025
Conference Paper
Status of GPU capabilities within the Shift Monte Carlo radiation transport code
Shift is a general-purpose Monte Carlo (MC) radiation transport code for fission, fusion, and national security applications. Shift has been adapted to efficiently run on GPUs in order to leverage leadership-class supercomputers. This work presents Shift’s current GPU capabilities. These include core radiation transport capabilities for eigenvalue and fixed-source simulati…
Sep, 2024
Journal
Assessing the Consequences of Postclosure Criticality in Spent Nuclear Fuel
The U.S. Department of Energy is funding research into studying the consequences of postclosure criticality on the performance of a generic repository by (1) identifying the features, events, and processes (FEPs) that need to be considered in such an analysis, (2) developing the tools needed to model the relevant FEPs in a postclosure performance assessment, and (3) conduc…
Dec, 2023
ORNL Report
Criticality Consequence Analysis Roadmap for Spent Nuclear Fuel Canister in a Repository
Dec, 2023
Journal
Deterministically estimated fission source distributions for Monte Carlo k-eigenvalue problems
The standard Monte Carlo (MC) k-eigenvalue algorithm involves iteratively converging the fission source distribution using a series of potentially time-consuming inactive cycles before quantities of interest can be tallied. One strategy for reducing the computational time requirements of these inactive cycles is the Sourcerer method, in which a deterministic eigenvalue cal…
Dec, 2023
Journal
Preliminary Validation of the Shift Monte Carlo Code for Fixed-Source Radiation Transport Problems
Simulations of neutron and photon transport using the Shift Monte Carlo radiation transport code are compared with experimental measurements and their corresponding benchmark simulations from several sets of experiments. Overall, Shift results match the calculations made by the benchmark teams quite well and match the measured values, which typically have large uncertainti…
Dec, 2023
Conference Paper
Source Range Detector Response Modeling using VERA
The CASL reactor simulation package VERA has been adapted to provide high-fidelity simulation capabilities for modeling source range detector response during subcritical reactor configurations. New features include the activation and shuffling of secondary source assemblies, transfer of burned fuel neutron emission data from the ORIGEN depletion solver to the MPACT determi…
Dec, 2023
Journal
Singular value decomposition of adjoint flux distributions for Monte Carlo variance reduction
Monte Carlo (MC) shielding calculations often use weight windows (WWs) and biased sources formed from a deterministic estimate of the adjoint flux to improve the convergence rate of tallies. This requires a significant amount of computer memory, which can limit the memory available for high-resolution tally output. A new method is proposed for reducing these memory require…
Dec, 2023
Conference Paper
Development of Terrenus, a Multiphysics Code for Spent Nuclear Fuel Cask Criticality Analysis