Scott Palmtag

Comparison of Generation of Higher-Order Neutron Scattering Cross Sections

Simulation of the BEAVRS benchmark using VERA

ANNALS OF NUCLEAR ENERGY, 145.

Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification

ANNALS OF NUCLEAR ENERGY, 132, 1–23.

Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR

VERA Core Simulator methodology for pressurized water reactor cycle depletion

In Nuclear Science and Engineering.

Assessment of Thermal Hydraulic Feedback Models

CTF Validation and Verification Manual

Demonstrate VERA Core Simulator Performance Improvements L2:PHI.P13.03

MPACT Standard Input User s Manual, Version 2.2.0

MPACT Theory Manual, Version 2.2.0

MPACT VERA Input User s Manual, Version 2.2.0

Simulation of CRUD Induced Power Shift using the VERA Core Simulator and MAMBA

VERA Benchmarking Results for Watts Bar Nuclear Plant Unit 1 Cycles 1-12

Development of COBRA-TF for Modeling Full-Core Reactor Operating Cycles

L3.PHI.CTF.P10.02-rev2 Coupling of Subchannel T/H (CTF) and CRUD Chemistry (MAMBA1D)

VERA Core Simulator Methodology for PWR Cycle Depletion

Validation and Application of the 3D Neutron Transport MPACT within CASL VERA-CS

An approach for coupled-code multiphysics core simulations from a common input

In Annals of Nuclear Energy (Oxford).

Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS

Improvements, enhancements, and optimizations of COBRA-TF

International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2013, 3, 2170–2181.

Validation of the lattice physics code LANCER02 with ENDF/B-VII library

International Conference on the Physics of Reactors 2010, PHYSOR 2010, 2, 1130–1139.

The advanced BWR core simulator AETNA02

International Conference on the Physics of Reactors 2008, PHYSOR 08, 2, 1282–1289.

Steady-state nuclear analysis methods at global nuclear fuel, invited

Transactions of the American Nuclear Society, 97, 557–558.

CASMO-4 and multigroup MCNP comparisons for MOX fuel assemblies

Transactions of the American Nuclear Society, 92, 508–509.

Microscopic depletion model in SIMULATE-4

Transactions of the American Nuclear Society, 92, 635–636.

A two-dimensional, semi-analytic expansion method for nodal calculations

MCNP: Multigroup/adjoint capabilities

(Vol. 4).

Measurements of actinide transmutation in the hard spectrum of a fast reactor