Dan Kotlyar

Assistant Professor

Georgia Institute of Technology

United States

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Dan Kotlyar is an Assistant Professor at the Georgia Institute of Technology in the United States. His research areas include nuclear thermal propulsion, computational methods in reactor analysis, and thermal hydraulics. He has authored several recent articles focusing on topics such as Monte Carlo methods for diffusion coefficients, multiphysics analysis of nuclear reactors, and energy-multigroup hybrid depletion schemes.

Articles (15)

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2025

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2023

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2020

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2020

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2020

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2020

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2020

Collaborators (3)

Ben Lindley

Associate Professor

University of Wisconsin-Madison

UNITED STATES

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Massimiliano Fratoni

Associate Professor

University of California, Berkeley

UNITED STATES

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Stefano Terlizzi

John and Jean Brennan Early Career Professor

Pennsylvania State University

UNITED STATES

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Dan Kotlyar

Research Interests

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Articles (15)

Validation and Verification of Serpent-Griffin Computational Sequence Using the SNAP 8 Experimental Reactor Dry Experiments

Consistent Method for Calculating Directional Diffusion Coefficients Using Monte Carlo

Preliminary Modeling of Nuclear Thermal Propulsion Cores Using Nodal Diffusion Codes with Jacobian-Free Newton Krylov–Generated Discontinuity Factors

Trade-Off Studies of a Radiantly Integrated TPV-Microreactor (RITMS) Design

Efficient continuous Energy-Multigroup hybrid depletion scheme using the Shift Monte Carlo code. Part I: Energy condensation sensitivity analysis

Integrated Steady-State System Package for Nuclear Thermal Propulsion Analysis Using Multi-Dimensional Thermal Hydraulics and Dimensionless Turbopump Treatment

Time dependent full-core multiphysics analysis of nuclear thermal propulsion reactors

Full-Core Coupled Neutronic, Thermal-Hydraulic, and Thermo-Mechanical Analysis of Low-Enriched Uranium Nuclear Thermal Propulsion Reactors

A density functional theory and neutron diffraction study of the ambient condition properties of sub-stoichiometric yttrium hydride

High-resolution thermal analysis of nuclear thermal propulsion fuel element using OpenFOAM

serpentTools: A Python Package for Expediting Analysis with Serpent

Low enriched nuclear thermal propulsion neutronic, thermal hydraulic, and system design space analysis

High temperature ultra-small modular reactor: Pre-conceptual design

On-The-Fly Prediction of Macroscopic Cross-Section Spatial Response to TH Perturbations Through Transfer Functions: Theory and First Results

A Transport-Free Method for Predicting the Post-Depletion Spatial Neutron Flux Distribution

Collaborators (3)

Ben Lindley

Massimiliano Fratoni

Stefano Terlizzi

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