@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={Abstract 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)â€“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 P3 (SP3) 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.}, 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}, abstractNote={In neutron transport calculations, it is common to specify material compositions in terms of constituent isotopes. Material compositions may be described by isotope number densities and associated microscopic cross sections. For general reaction cross sections, the macroscopic cross sections of a composition are simply the summation of the sum of the products of isotope number densities and microscopic cross sections. A notable exception is the mixture of the neutron fission spectrum in fissile material. To demonstrate proper and improper mixture of the neutron fission spectrum, a zero-dimensional test problem is developed. Using the test problem, it is demonstrated that the improper mixing of the fission spectrum results in an eigenvalue error of approximately 65 pcm. Though this may seem small, any error in such a simple calculation is unacceptable. In similar infinite-homogeneous test problems, eigenvalue errors of more than 300 pcm have been observed. It is also shown that the error due to fission spectrum mixing can be exaggerated for computer program verification purposes. It is concluded that the proper mixture of the neutron fission spectrum is essential for accurate neutron transport simulations. The effect of improper neutron fission spectrum mixture is demonstrated and quantified and this test problem may be used for verification purposes in the future.}, 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}, abstractNote={The generation of high scattering order neutron scattering cross sections consistent with high-fidelity simulations remains an area of active research. Popular options include generating cross sections from continuous energy Monte Carlo calculations or from a deterministic neutron transport calculation with high-fidelity tabulated cross sections. Both options present challenges. Monte Carlo simulations can naturally process continuous energy cross-section data and allow for the general description of anisotropic neutron scattering given a scattering law. However, Monte Carlo simulations are inherently related to particle weighting and it has been suggested that this may be unacceptable for generating high-order neutron scattering cross sections. Deterministic neutron transport calculations can easily calculate high-order moments of the flux to appropriately calculate higher-order neutron scattering cross sections but are generally limited by discretization of space, energy, and angle. In this work, the trade-offs between generation of high-order neutron scattering cross sections via Monte Carlo and deterministic neutron transport methods are investigated. The methods implemented in the Monte Carlo computer program Serpent 2 and the deterministic fast reactor neutron cross section generator MC2-3 are compared. Cross sections resulting from these methods are used in Rattlesnake, a deterministic neutron transport code developed by Idaho National Laboratory, and results are compared to a reference continuous energy Monte Carlo calculation. Whereas previous work investigating the effects of anisotropic neutron scattering has focused on light water reactor simulations, this work focuses on high-order neutron scattering cross sections as they relate to fast reactor simulations. To investigate the consequences of the Serpent 2 and MC2-3 methodologies, a test problem is developed. The test problem is a one-dimensional geometry with fast reactor materials designed to demonstrate deep penetration and exacerbate the effects of high-order neutron scattering. Based on the results of the deep penetration test problem, it is concluded that P3 neutron scattering cross sections are sufficient to describe anisotropic scattering in fast reactor materials. Any neutron scattering of order higher than P3 offers negligible change in the eigenvalue of the test problem. Additionally, it is determined that the methodology as implemented in Serpent 2 is applicable for generating high-order neutron scattering cross sections through at least P3 in fast reactor materials.}, 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} }