Scott Palmtag

A Design and Optimization Methodology for Liquid Metal Fast Reactors

INTERNATIONAL JOURNAL OF ENERGY RESEARCH, 2023.

Fuel cycle cost comparison between lead and sodium cooled fast reactors

Nuclear Engineering and Design.

Solving the Neutron Transport Equation for Microreactor Modeling Using Unstructured Meshes and Exascale Computing Architectures

Dawn, W. C., & Palmtag, S. (2023, May 11). NUCLEAR SCIENCE AND ENGINEERING, Vol. 5.

CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors

NUCLEAR ENGINEERING AND DESIGN, 397.

Control rod modeling in liquid metal-cooled fast reactors

Kiefer, T. M., Dawn, W. C., Al-Dawood, K., & Palmtag, S. (2022, July). https://www.osti.gov/biblio/23178744

LMFR design and optimization methodology

Al-Dawood, K. A., & Palmtag, S. P. (2022, July). https://www.osti.gov/biblio/23178727

VERA BWR Progression Problems

Lawing, C. ; Palmtag, S. ; & Asgari, M. (2022, May). https://www.osti.gov/biblio/1890346

A MULTIPHYSICS SIMULATION SUITE FOR SODIUM COOLED FAST REACTORS

EPJ Web of Conferences, 247, 06019.

A multiphysics simulation suite for liquid metal-cooled fast reactors

ANNALS OF NUCLEAR ENERGY, 159.

Analysis of Approximations in Modeling of BWR Bundle Void Distributions

BWR Progression Problems

QUANTIFICATION OF NUCLEAR DATA AND MANUFACTURING UNCERTAINTIES IN VERA

EPJ Web of Conferences, 247, 15014.

Quantum Annealing Stencils with Applications to Fuel Loading of a Nuclear Reactor

2021 IEEE INTERNATIONAL CONFERENCE ON QUANTUM COMPUTING AND ENGINEERING (QCE 2021) / QUANTUM WEEK 2021, pp. 265–275.

Secondary-Source Core Reload Modeling with VERA

NUCLEAR SCIENCE AND ENGINEERING, 195(3), 320–337.

VERA-CS VALIDATION OF CRITICAL EXPERIMENTS

EPJ Web of Conferences, 247, 10014.

CTF Validation and Verification

Comparison of Generation of Higher-Order Neutron Scattering Cross Sections

In Idaho National Laboratory.

Design of an Economical Low-Power Fast Spectrum Molten Chloride Reactor

Transactions of the American Nuclear Society - Volume 123.

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

https://www.osti.gov/scitech/biblio/1325432-assessment-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

https://www.osti.gov/scitech/biblio/1325433-simulation-crud-induced-power-shift-using-vera-core-simulator-mamba

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

https://www.osti.gov/scitech/biblio/1260079-vera-benchmarking-results-watts-bar-nuclear-plant-unit-cycles

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

https://www.osti.gov/scitech/biblio/1185922-development-cobra-tf-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

https://www.osti.gov/scitech/biblio/1215576-vera-core-simulator-methodology-pwr-cycle-depletion

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

https://www.osti.gov/scitech/biblio/1214019-validation-application-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

https://www.osti.gov/scitech/biblio/1154814-coupled-neutronics-thermal-hydraulic-solution-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. http://www.scopus.com/inward/record.url?eid=2-s2.0-84883365179&partnerID=MN8TOARS

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. http://www.scopus.com/inward/record.url?eid=2-s2.0-79952378953&partnerID=MN8TOARS

The advanced BWR core simulator AETNA02

International Conference on the Physics of Reactors 2008, PHYSOR 08, 2, 1282–1289. http://www.scopus.com/inward/record.url?eid=2-s2.0-79953837410&partnerID=MN8TOARS

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

Transactions of the American Nuclear Society, 97, 557–558. http://www.scopus.com/inward/record.url?eid=2-s2.0-55349115707&partnerID=MN8TOARS

CASMO-4 and multigroup MCNP comparisons for MOX fuel assemblies

Transactions of the American Nuclear Society, 92, 508–509. http://www.scopus.com/inward/record.url?eid=2-s2.0-33244479401&partnerID=MN8TOARS

Microscopic depletion model in SIMULATE-4

Transactions of the American Nuclear Society, 92, 635–636. http://www.scopus.com/inward/record.url?eid=2-s2.0-33244493781&partnerID=MN8TOARS

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

https://www.osti.gov/scitech/biblio/204716-measurements-actinide-transmutation-hard-spectrum-fast-reactor