Adam Dix
Engineering · Purdue University West Lafayette
Publications
25
Citations
53
Est. group size
~8
Recurring co-author estimate
Active years
19
Publishing since 2008
Adam Dix works on nuclear engineering, focusing on the thermal and fluid behavior of advanced reactor designs. His research includes experiments and modeling of two-phase flows (mixtures of gas bubbles and liquid, such as air and water), pressure drop and friction in fuel rod bundles used in sodium-cooled fast reactors, and computer code validation for simulating reactor coolant flow. This work supports the design and safety analysis of next-generation nuclear reactors, including sodium fast reactors and the Versatile Test Reactor.
Publication output has grown over the last decade, rising from sporadic single papers in 2017-2019 to a more steady output of roughly 4-6 papers per year since 2020.
Generated by claude-sonnet-5 from public bibliographic data · Jul 20, 2026
- Experimental Study on Air-Water Two-Phase Bubbly Flow Across a Vertical U-Bend
Nuclear Technology · 2026
- Relative Velocity in Horizontal Bubbly Two-Phase Flows: Experiments and Modeling
Nuclear Technology · 2026
- Investigation of Relative Velocity in Non-Vertical Air-Water Bubbly Two-Phase Flows
SSRN Electronic Journal · 2025
- Investigation of relative velocity in non-vertical air-water bubbly two-phase flows
International Journal of Multiphase Flow · 2025
- Code Validation of SAM Using Forced and Natural-Circulation Data from NACIE-UP Benchmark
2023
- Summary of a Novel Friction Factor Model for Wire-Wrapped Rod Bundle
Transactions of the American Nuclear Society · 2023
- Development of a Sodium Fast Reactor Cartridge Loop Testing Capability for the Versatile Test Reactor
Nuclear Science and Engineering · 2022
- A novel friction factor model for wire-wrapped rod bundles
Nuclear Engineering and Design · 2022
- Pressure Drop in Seven-Pin Wire-Wrapped Rod Bundle for the Sodium Cartridge Loop in Versatile Test Reactor
Nuclear Science and Engineering · 2022
- Evaluation of Pressure Drop Models for 7-Pin Wire-Wrapped Rod Bundles
Transactions of the American Nuclear Society · 2022
- CFD Simulations of the Sodium Cartridge Loop of the Versatile Test Reactor
Transactions of the American Nuclear Society - Volume 123 · 2020
- CARLITA Code Benchmarking with SAM for LBE Natural Circulation Flow
Transactions of the American Nuclear Society · 2020
- Development of Integrated Mechanistic Source Term Assessment Software for Liquid Metal Reactors
Transactions of the American Nuclear Society · 2020
- Transactions of the American Nuclear Society×10
- Nuclear Technology×3
- Nuclear Engineering and Design×2
- Nuclear Science and Engineering×2
- Geoenergy Science and Engineering×1
- S M Kim
Engineering · Purdue University West Lafayette
- Mamoru Ishii
Engineering · Purdue University West Lafayette
- Zhuoran Dang
Engineering · Purdue University West Lafayette
- Seungjin Kim
Engineering · Purdue University West Lafayette
- Mamoru Ishii
Engineering · Purdue University West Lafayette
This profile was generated automatically from public scholarly data (OpenAlex). Group size and activity levels are estimates derived from co-authorship patterns.
Last updated Jul 20, 2026.
Claim or correct this profile