@article{dawn_palmtag_2023, title={Solving the Neutron Transport Equation for Microreactor Modeling Using Unstructured Meshes and Exascale Computing Architectures}, volume={5}, ISSN={["1943-748X"]}, url={https://doi.org/10.1080/00295639.2023.2189510}, DOI={10.1080/00295639.2023.2189510}, abstractNote={The Microreactor Exascale eZ CALculation (MEZCAL) tool has been developed to accurately and efficiently solve the neutron transport equation in general, unstructured meshes to support the design and modeling of microreactors. MEZCAL solves the self-adjoint angular flux form of the neutron transport equation using the finite element method. As the neutron transport equation is computationally expensive to solve, MEZCAL is designed to efficiently use exascale computing architectures, with an emphasis on graphics processing unit computing. To leverage existing tools, MEZCAL is built using the MFEM library and uses solvers from HYPRE, PETSc, and SLEPc. Verification of the neutron transport solver in MEZCAL is demonstrated with the solution to a one-dimensional cylindrical problem that has a semi-analytic solution. After verification, a realistic microreactor based on the MARVEL microreactor design is modeled using MEZCAL. Spatial and angular refinement results are presented for a two-dimensional model of the MARVEL microreactor, and the eigenvalue is converged to approximately 60 pcm. This convergence required a very fine mesh and more than 3.76 Billion Degrees Of Freedom (BDOF). Preliminary results are also presented for a three-dimensional model of the MARVEL microreactor. Finally, a weak scaling study is performed to investigate how the methods in MEZCAL will scale for larger problems with the next generation of exascale computing architectures.}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Dawn, William C. and Palmtag, Scott}, year={2023}, month={May} }
 @article{salko_wysocki_blyth_toptan_hu_kumar_dances_dawn_sung_kucukboyaci_et al._2022, title={CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors}, volume={397}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2022.111927}, abstractNote={CTF is a thermal hydraulic (T/H) subchannel tool that has been extensively developed over the past ten years as part of the Consortium for Advanced Simulation of Light Water Reactors (CASL) program. The code was selected early in the CASL program for support of high-impact challenge problems that were found to be relevant to the nuclear industry and its currently operating fleet of pressurized water reactors (PWRs), including issues such as departure from nucleate boiling (DNB), crud-induced power shifts (CIPSs), and reactivity-insertion accidents (RIAs). By incorporating CTF into the multiphysics Virtual Environment for Reactor Application (VERA) core simulator software developed by CASL, CTF has become the primary means of providing fluid and fuel thermal feedback, as well as T/H figure-of-merits (FOMs) in large-scale reactor simulations. With the goal of solving industry challenge problems, CASL placed great emphasis on developing high-quality, high-performance, validated software tools that offer higher fidelity than what is currently possible with current industry methods. In support of this effort, CTF was developed from a research tool into an nuclear quality assurance (NQA-1)–compliant, production-level software tool that is capable of addressing the stated challenge problems and goals of CASL. This paper presents a review of the major technological achievements that were realized in developing CTF over the past decade of the CASL program and presents an overview of the code solution approach and closure models.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Salko, Robert and Wysocki, Aaron and Blyth, Taylor and Toptan, Aysenur and Hu, Jianwei and Kumar, Vineet and Dances, Chris and Dawn, William and Sung, Yixing and Kucukboyaci, Vefa and et al.}, year={2022}, month={Oct} }
 @article{dawn_palmtag_2021, title={A Multiphysics Simulation Suite for Sodium Cooled Fast Reactors}, volume={247}, url={http://dx.doi.org/10.1051/epjconf/202124706019}, DOI={10.1051/epjconf/202124706019}, abstractNote={A simulation suite has been developed to model reactor power distribution and multiphysics feedback effects in Sodium-cooled Fast Reactors (SFRs). This suite is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified P 3 (SP 3 ) neutron transport equations. In the FEM implementation, two-dimensional triangular elements and three-dimensional wedge elements are selected. Wedge elements are selected for their natural description of hexagonal geometry common to fast reactors. Thermal feedback effects within fast reactors are modeled within the simulation suite. A thermal hydraulic model is developed, modeling both axial heat convection and radial heat conduction within fuel assemblies. A thermal expansion model is included and is demonstrated to significantly affect reactivity. This simulation suite has been employed to model the Advanced Burner Reactor (ABR) benchmark, specifically the MET-1000. It has been demonstrated that these models sufficiently describe the multiphysics feedback phenomena and can be used to estimate multiphysics reactivity feedback coefficients.}, journal={EPJ Web of Conferences}, publisher={EDP Sciences}, author={Dawn, William C. and Palmtag, Scott}, editor={Margulis, M. and Blaise, P.Editors}, year={2021}, pages={06019} }
 @article{dawn_palmtag_2021, title={A multiphysics simulation suite for liquid metal-cooled fast reactors}, volume={159}, ISSN={["1873-2100"]}, url={http://dx.doi.org/10.1016/j.anucene.2021.108213}, DOI={10.1016/j.anucene.2021.108213}, abstractNote={A new multiphysics simulation suite has been created to model Liquid Metal-cooled Fast Reactors (LMFRs). LUPINE: “LMFR Utility for Physics Informed Nuclear Engineering” is based on the Finite Element Method (FEM) and employs a general, unstructured mesh to solve the Simplified PN (SPN) neutron transport equations and multiphysics models to simulate thermal hydraulics and thermal expansion. Results of validation benchmarks are presented using the SPN solver and runtime results indicate that the SPN solver scales well for increasing SPN truncation order. To demonstrate the multiphysics coupling capabilities of LUPINE, the Advanced Burner Reactor (ABR) MET-1000 benchmark is modeled using coupled neutronics, thermal hydraulics, and thermal expansion models. It is shown that the SP3 method is sufficient to model the SPN effects in the ABR. The multiphysics models are showcased by calculating several multiphysics reactivity coefficients including: power defect, thermal expansion coefficient, Doppler coefficient, and Coolant Temperature Coefficient (CTC).}, journal={ANNALS OF NUCLEAR ENERGY}, author={Dawn, William C. and Palmtag, Scott}, year={2021}, month={Sep} }
 @misc{systems and methods for airflow control in reactor passive decay heat removal_2021, url={https://patents.google.com/patent/US10937557B2}, journal={GE Hitachi Nuclear Energy Americas LLC}, year={2021}, month={Mar} }
 @book{dawn_2020, title={An Analytic Benchmark for the Solution to the Isotopic Fission Spectrum Mixture Problem}, url={http://dx.doi.org/10.2172/1593873}, DOI={10.2172/1593873}, journal={[]}, institution={Office of Scientific and Technical Information (OSTI)}, author={Dawn, William}, year={2020}, month={Jan} }
 @book{dawn_ortensi_dehart_palmtag_2020, title={Comparison of Generation of Higher-Order Neutron Scattering Cross Sections}, url={http://dx.doi.org/10.2172/1593864}, DOI={10.2172/1593864}, journal={[]}, institution={Office of Scientific and Technical Information (OSTI)}, author={Dawn, William and Ortensi, Javier and DeHart, Mark and Palmtag, Scott}, year={2020}, month={Jan} }
 @article{dawn_2019, title={Simulation of Fast Reactors with the Finite Element Method and Multiphysics Models}, url={http://www.lib.ncsu.edu/resolver/1840.20/36547}, journal={North Carolina State University}, publisher={North Carolina State University}, author={Dawn, William C.}, year={2019}, month={May} }
 @misc{loewen_sineath_molinaro_dawn_garcia_meek_day_2018, place={Wilmington, NC}, title={Acoustic Flowmeter and Methods of Using Same}, url={https://patents.google.com/patent/US20180277267A1/en?oq=20180277267}, number={20180277267}, journal={GE-Hitachi Nuclear Americas LLC}, author={Loewen, Eric P. and Sineath, James P. and Molinaro, Dean D. and Dawn, William C. and Garcia, William J. and Meek, Oscar L. and Day, Patrick K.}, year={2018}, month={Sep} }
 @misc{loewen_sineath_molinaro_dawn_sprague_marshall_melito_2018, place={Wilmington, NC}, title={Intermixing Feedwater Sparger Nozzles and Methods for Using the Same in Nuclear Reactors}, url={https://patents.google.com/patent/US20180277265A1}, number={20180277265}, journal={GE-Hitachi Nuclear Americas LLC}, author={Loewen, Eric P. and Sineath, James P. and Molinaro, Dean D. and Dawn, William C. and Sprague, Robin D. and Marshall, Theron D. and Melito, Joel P.}, year={2018}, month={Sep} }