@article{al-dawood_palmtag_2024, title={Advanced Liquid-Metal-Cooled Fast Reactor Core Design Using Modern Optimization Algorithms}, volume={5}, ISSN={["1943-748X"]}, url={https://doi.org/10.1080/00295639.2024.2347696}, DOI={10.1080/00295639.2024.2347696}, abstractNote={When performing core loading pattern design in fast spectrum reactors, it is often assumed that the larger neutron mean free path in fast reactors makes the core loading pattern less significant than in thermal reactors. Due to this assumption, the literature often includes homogeneous core designs for liquid-metal-cooled fast reactors (LMFRs). In this paper, heterogeneous loading patterns are investigated using modern LMFR multiphysics analysis. It was found that the figures of merit (FOMs) used to design cores are very sensitive to the core loading pattern, and better core designs (measured by the FOM) can be obtained from heterogeneous designs.}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Al-Dawood, Khaldoon and Palmtag, Scott}, year={2024}, month={May} }
 @article{al-dawood_palmtag_2024, title={METAL: Methodology for liquid metal fast reactor core economic design and fuel loading pattern optimization}, volume={173}, url={https://doi.org/10.1016/j.pnucene.2024.105232}, DOI={10.1016/j.pnucene.2024.105232}, abstractNote={The design of Liquid Metal-cooled Fast Reactor (LMFR) cores encompass two levels of design. The first is the physical core design, which is concerned with the design of the fuel assembly geometry in the context of a full core. The second is the fuel loading design, which is an in-core fuel management problem concerned with the design of fuel enrichment and core loading pattern. In this paper, an optimized fuel design search is developed to improve core loading patterns. As part of the implementation, the multiphysics simulation suite LUPINE has been enhanced to allow for fuel shuffles and an in-core fuel management optimization methodology has been added with the objective of reducing the fuel cost. The design methodology is referred to as Methodology for Economical opTimization of Applied LMFR (METAL), and this methodology is demonstrated using the popular Super Power Reactor Innovative Small Modular (SPRISM) sodium-cooled fast reactor (SFR) as a reference core. One challenge currently facing the deployment of LMFRs is the availability of TRansUranium (TRU) and high assay low enriched uranium (HALEU) driver fuel. The two forms of fuel are expensive and not widely available, which poses a challenge on the startup of LMFR cores. To address the availability of driver fuel, and to lower fuel costs, low enriched uranium (LEU) blankets are investigated as replacements to the natural uranium or depleted uranium blankets typically used. The advantage of this solution is the wide availability of LEU and its ability to lower the amount of TRU or HALEU driver fuel needed. The design objectives, constraints, and optimization algorithm for a SFR are identified. Then, METAL is applied to design a SFR core fueled by TRU driver fuel assemblies and LEU blanket assemblies. To demonstrate the advantage of introducing LEU blankets, METAL is also used to design another SFR core following the same objectives and constraints, but using naturally enriched blankets. By comparing the two cores, it was found that the introduction of LEU blankets results in a ∼28% reduction in the driver fuel mass requirements. Upon development of a levelized fuel cycle cost (LFCC) model, the 28% reduction in driver fuel mass corresponds to a 10% decrease in the LFCC. Additional advantages of using LEU blankets include reducing the core conversion ratio (CR), and improving the assembly radial power peaking factor (RPPF), which enhances the safety and non-proliferation performances of the SFR core.}, journal={Progress in Nuclear Energy}, author={Al-Dawood, Khaldoon and Palmtag, Scott}, year={2024}, month={Aug} }
 @article{al-dawood_palmtag_2023, title={A Design and Optimization Methodology for Liquid Metal Fast Reactors}, volume={2023}, ISSN={["1099-114X"]}, url={https://doi.org/10.1155/2023/6846467}, DOI={10.1155/2023/6846467}, abstractNote={A liquid metal fast reactor (LMFR) design and optimization methodology (DOM) has been developed. The methodology effectively explores a search space by initially sampling the search space, excluding invalid design samples prior to performing expensive multiphysics analysis, and then performing local searches of the design space. The design samples are evaluated using the multiphysics capabilities of the LUPINE LMFR simulation suite. Two studies have been performed to demonstrate DOM. First, the Westinghouse long-life core lead fast reactor (WLFR) is optimized. This reactor is 950 
                     
                        
                           
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                   and fueled with uranium nitride (UN) fuel which has a natural nitrogen isotopic abundance. The objective of the optimization is the reduction of the levelized fuel cycle cost (LFCC) while complying with the design constraints. Considering the challenges associated with using natural nitrogen in nitride fuel, a second study was performed to design a competitive 15N-enriched UN-fueled long-life core LFR. Based on this design, the cost of the 15N enrichment process necessary to achieve a competitive LFCC was calculated.}, journal={INTERNATIONAL JOURNAL OF ENERGY RESEARCH}, author={Al-Dawood, Khaldoon and Palmtag, Scott}, editor={Kuzmin, AndreyEditor}, year={2023}, month={Mar} }
 @article{al-dawood_palmtag_2023, title={Fuel cycle cost comparison between lead and sodium cooled fast reactors}, volume={414}, ISSN={["1872-759X"]}, url={https://doi.org/10.1016/j.nucengdes.2023.112583}, DOI={10.1016/j.nucengdes.2023.112583}, abstractNote={Sodium and lead coolants are the most recognized coolants for Liquid Metal-cooled Fast Reactor (LMFR). The two coolants have significantly different physical, thermal and chemical properties, which results in major core design differences between lead- and sodium-cooled systems. This work attempts to quantify the impact of the coolant type on the fuel cycle cost of a LMFR by comparing the Levelized Fuel Cycle Cost (LFCC). The paper is composed of two studies with the goal of comparing the advantage of the superior neutronic performance of lead versus the advantage of the tight lattice that can be achieved with sodium coolant. In the first study, the influence of the neutronic differences between lead and sodium on the fuel cycle cost are isolated and quantified. This is achieved by comparing the fuel cycle performance of a Lead-cooled Fast Reactor (LFR) core model to that of a Non-optimized SFR (NSFR) model. The two models have a similar geometry, fuel loading, fuel type, and structural materials. The only major difference between the two cores was the coolant (i.e. lead vs. sodium). Based on this comparison, it is shown that the better neutronic performance of lead coolant results in longer cycle length for the LFR core. It is also shown that this difference in cycle length can result in up to 30.8% better LFCC for a lead-cooled system compared to a sodium-cooled one. In the second study, a Sodium-cooled Fast Reactor (SFR) is designed that is optimized for the sodium coolant. The design process for the Long-life core Sodium Fast Reactor (LSFR) realizes the compatibility of sodium coolant with structural materials and allows for a more compact design. The design parameters and constraints of the design are presented, and the objective of the design is to reduce the LFCC. In this study, the design with the sodium coolant offers a 2% reduction in the LFCC, along with a smaller core size. The overall results show that if you compare sodium and lead using the same reactor geometry, the neutronic benefits of lead coolant can lead to a 30% reduction in the LFCC. However, if you optimize the geometry to take advantage of the higher power density allowed with sodium coolant, the sodium coolant can offer a 2% LFCC reduction or a smaller core size.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Al-Dawood, Khaldoon and Palmtag, Scott}, year={2023}, month={Dec} }
 @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)–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} }
 @inproceedings{kiefer_dawn_al-dawood_palmtag_2022, title={Control rod modeling in liquid metal-cooled fast reactors}, url={https://www.osti.gov/biblio/23178744}, author={Kiefer, T. M.; and Dawn, W. C.; and Al-Dawood, K.; and Palmtag, S.}, year={2022}, month={Jul} }
 @inproceedings{al-dawood_palmtag_2022, title={LMFR design and optimization methodology}, url={https://www.osti.gov/biblio/23178727}, author={Al-Dawood, Khaldoon A.; and Palmtag, Scott P.}, year={2022}, month={Jul} }
 @inproceedings{lawing_palmtag_asgari_2022, title={VERA BWR Progression Problems}, url={https://www.osti.gov/biblio/1890346}, author={Lawing, Chase ; and Palmtag, Scott ; and Asgari, Mehdi}, year={2022}, month={May} }
 @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}, month={Feb}, 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} }
 @book{lawing_palmtag_kropaczek_2021, title={Analysis of Approximations in Modeling of BWR Bundle Void Distributions}, url={https://www.osti.gov/biblio/1845788}, DOI={10.2172/1845788}, abstractNote={trends agreed with previous diabatic measurements in which the center subchannel had the highest quality and mass flux, while the corner subchannel had the lowest.}, author={Lawing, Chase ; and Palmtag, Scott ; and Kropaczek, Dave J.}, year={2021}, month={Aug} }
 @book{lawing_palmtag_asgari_2021, title={BWR Progression Problems}, url={https://www.osti.gov/biblio/1838995}, DOI={10.2172/1838995}, abstractNote={Under the Consortium for the Advanced Simulation of Light Water Reactors (CASL) Program, the Virtual Environment for Reactor Applications (VERA) was developed with the primary focus to model pressurized water reactors (PWRs). Recently, a new project was started to extend the modeling capability in VERA to model boiling water reactors (BWRs). The new project is called “Modeling and Analysis of Exelon BWRs for Eigenvalue and Thermal Limits Predictability,” and it is led by Oak Ridge National Laboratory (ORNL) and Exelon with participation from Global Nuclear Fuel and three universities: North Carolina State University (NCSU), the University of Michigan, and the University of Illinois.}, author={Lawing, Chase ; and Palmtag, Scott ; and Asgari, Mehdi}, year={2021}, month={Sep} }
 @article{sedota_palmtag_2021, title={QUANTIFICATION OF NUCLEAR DATA AND MANUFACTURING UNCERTAINTIES IN VERA}, volume={247}, DOI={10.1051/epjconf/202124715014}, abstractNote={Uncertainty quantification (UQ) was performed using the Consortium for the Advanced Simulation of Light Water Reactors (CASL) multiphysics core simulator VERA. Typically, only nuclear data cross sections are considered when trying to obtain uncertainty information in reactor simulation applications. In this paper, uncertainty in both nuclear cross section data and fuel manufacturing processes is analyzed using VERA. Uncertainties due to cross sections was determined by generating one thousand perturbed cross section libraries using the cross section covariance data provided in the evaluated nuclear data library. Uncertainty due to manufacturing was also determined using stochastic sampling and VERA. The use of similar stochastic sampling techniques for the same problems allows for the direct comparison of uncertainty stemming from the two sources of uncertainty. Sample size is considered due to the potentially large computational cost of stochastic sampling techniques, as is demonstrated in a full core depletion. It was found that for the Pressurized Water Reactor (PWR) pincell case at Hot Zero Power (HZP), the standard deviation in the neutron multiplication factor produced by material uncertainty was 101 pcm, while the standard deviation in the neutron multiplication factor produced by cross section uncertainty was 730 pcm. While the uncertainty in neutron multiplication factor due to cross section uncertainty is larger than uncertainty due to manufacturing uncertainties, neglecting manufacturing uncertainties may be an unacceptable oversight in certain high-precision simulation applications.}, journal={EPJ Web of Conferences}, publisher={EDP Sciences}, author={Sedota, Christopher and Palmtag, Scott}, editor={Margulis, M. and Blaise, P.Editors}, year={2021}, month={Feb}, pages={15014} }
 @article{fustero_palmtag_mueller_2021, title={Quantum Annealing Stencils with Applications to Fuel Loading of a Nuclear Reactor}, DOI={10.1109/QCE52317.2021.00044}, abstractNote={A method for mapping quadratic unconstrained binary optimizations expressed as nearest neighbor stencils onto contemporary quantum annealing machines is developed. The method is shown to be scalable in providing higher utilization of annealing hardware resources than prior work. Applying the technique to the problem of determining an effective fuel loading pattern for nuclear reactors shows that densely mapped quantum stencils result in higher fidelity solutions of optimization problems then the sparser default solutions. These results are likely to generalize to quadratic unconstrained binary optimizations that can be expressed as dense quantum stencils, thereby improving optimization results obtained from noisy quantum devices.}, journal={2021 IEEE INTERNATIONAL CONFERENCE ON QUANTUM COMPUTING AND ENGINEERING (QCE 2021) / QUANTUM WEEK 2021}, author={Fustero, Joseph and Palmtag, Scott and Mueller, Frank}, year={2021}, pages={265–275} }
 @article{gentry_collins_davidson_davidson_evans_godfrey_hart_ilas_johnson_kim_et al._2021, title={Secondary-Source Core Reload Modeling with VERA}, volume={195}, ISSN={["1943-748X"]}, DOI={10.1080/00295639.2020.1820797}, abstractNote={Abstract The CASL reactor simulation package VERA has been adapted to provide high-fidelity simulation capabilities for modeling source range detector response during subcritical reactor configurations. New features include the activation and shuffling of secondary-source assemblies, use of burned fuel neutron emission data from the ORIGEN depletion solver to the MPACT deterministic neutron transport solver, allowance of user-defined sources in MPACT based on material composition, ability to solve the subcritical source-driven system with neutron multiplication using the MPACT diffusion solver, and transfer of the calculated fission source from MPACT to the continuous-energy Monte Carlo solver Shift for final detector response evaluation using the CADIS methodology for variance reduction. These new capabilities were benchmarked against Watts Bar Unit 1 plant operating data for the first few fuel loading steps and were found to demonstrate excellent agreement with the measured data.}, number={3}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Gentry, Cole and Collins, Benjamin and Davidson, Eva and Davidson, Gregory and Evans, Thomas and Godfrey, Andrew and Hart, Shane and Ilas, Germina and Johnson, Seth and Kim, Kang Seog and et al.}, year={2021}, month={Mar}, pages={320–337} }
 @article{novellino_palmtag_2021, title={VERA-CS VALIDATION OF CRITICAL EXPERIMENTS}, volume={247}, DOI={10.1051/epjconf/202124710014}, abstractNote={VERA is a suite of multiphysics codes which uses MPACT to model neutron transport in light water reactors (LWRs) [1]. In this paper, we validate MPACT by modeling critical experiments conducted at the IPEN/MB-01 and B&W facilities. We modeled critical loading experiments with a variety of different fuel pins and materials placed in the core. The experiments were modeled in two dimensions using MPACT and an axial buckling term. Default mesh parameters exist in MPACT for modeling larger reactor cores, and a mesh convergence study was performed to find appropriate mesh parameters for modeling the smaller critical reactors. The keff results show a consistent bias and small standard deviation for the IPEN/MB-01 reactor and a small bias and small standard deviation for the B&W facility. Overall, the results show that MPACT performs well for modeling small critical reactors.}, journal={EPJ Web of Conferences}, publisher={EDP Sciences}, author={Novellino, Vincent and Palmtag, Scott}, editor={Margulis, M. and Blaise, P.Editors}, year={2021}, month={Feb}, pages={10014} }
 @book{salko_wysocki_toptan_porter_zhao_blyth_magedanz_dances_gorgar_gosdin_et al._2020, title={CTF Validation and Verification}, DOI={10.2172/1731045}, abstractNote={Coolant-Boiling in Rod Arrays- Two Fluids (COBRA-TF) is a thermal/hydraulic (T/H) simulation code designed for Light Water Reactor (LWR) analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of nine conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and has been used and modified by several institutions over the last several decades. COBRA-TF is also used at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Modeling Group (RDFMG) and has been improved, updated, and subsequently became the PSU RDFMG version of COBRA-TF (CTF). One part of the improvement process includes validating the methods in CTF. This document seeks to provide a certain level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model-rod bundle geometries with operating conditions that are representative of prototypical Pressurized Water Reactor (PWR)s and Boiling Water Reactor (BWR)s in both normal and accident conditions. This is done by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest.}, institution={Office of Scientific and Technical Information (OSTI)}, author={Salko, Robert, Jr and Wysocki, Aaron and Toptan, A. and Porter, Nathan and Zhao, Xingang and Blyth, Taylor and Magedanz, Jeffrey and Dances, Christopher and Gorgar, M. and Gosdin, Chris and et al.}, year={2020}, month={Nov} }
 @book{dawn_ortensi_dehart_palmtag_2020, title={Comparison of Generation of Higher-Order Neutron Scattering Cross Sections}, url={https://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={Idaho National Laboratory}, 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{verrico_palmtag_thompson_novellino_goodman_2020, title={Design of an Economical Low-Power Fast Spectrum Molten Chloride Reactor}, DOI={10.13182/t123-32905}, journal={Transactions of the American Nuclear Society - Volume 123}, publisher={AMNS}, author={Verrico, L. and Palmtag, S. and Thompson, T. and Novellino, V. and Goodman, C.}, year={2020} }
 @article{collins_godfrey_stimpson_palmtag_2020, title={Simulation of the BEAVRS benchmark using VERA}, volume={145}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2020.107602}, journal={ANNALS OF NUCLEAR ENERGY}, author={Collins, Benjamin and Godfrey, Andrew and Stimpson, Shane and Palmtag, Scott}, year={2020}, month={Sep} }
 @article{kim_gentry_godfrey_liu_palmtag_2019, title={Development of the multigroup cross section library for the CASL neutronics simulator MPACT: Verification}, volume={132}, ISSN={["0306-4549"]}, url={https://doi.org/10.1016/j.anucene.2019.03.041}, DOI={10.1016/j.anucene.2019.03.041}, abstractNote={The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3D whole core transport code being developed for the CASL toolset, known as the Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include a subgroup method for resonance self-shielding and a whole-core transport solver with a 2D/1D synthesis method. Depletion capability is supported by coupling SCALE-ORIGEN with MPACT. A new ENDF/B-VII.1 MPACT 51-group library has been developed by using the AMPX/SCALE and CASL-XSTools code packages based on the procedure introduced in the companion paper, “Development of the Multigroup Cross Section Library for the CASL Neutronics Simulator MPACT: Method and Procedure.” The cross section library generation procedure, cross section library, and MPACT transport capability have been verified by performing various benchmark calculations and comparing the benchmark results to the continuous-energy (CE) Monte Carlo results. The code-to-code comparison for the benchmarks shows that the CASL neutronics simulator MPACT and its 51-group library are working reasonably.}, journal={ANNALS OF NUCLEAR ENERGY}, publisher={Elsevier BV}, author={Kim, Kang Seog and Gentry, Cole A. and Godfrey, Andrew T. and Liu, Yuxuan and Palmtag, Scott}, year={2019}, month={Oct}, pages={1–23} }
 @book{kim_clarno_gentry_wiarda_williams_kochunas_liu_palmtag_godfrey_2017, title={Development of the V4.2m5 and V5.0m0 Multigroup Cross Section Libraries for MPACT for PWR and BWR}, url={https://www.osti.gov/scitech/biblio/1352789-development-v4-v5-multigroup-cross-section-libraries-mpact-pwr-bwr}, DOI={10.2172/1352789}, abstractNote={The MPACT neutronics module of the Consortium for Advanced Simulation of Light Water Reactors (CASL) core simulator is a 3-D whole core transport code being developed for the CASL toolset, Virtual Environment for Reactor Analysis (VERA). Key characteristics of the MPACT code include (1) a subgroup method for resonance selfshielding and (2) a whole-core transport solver with a 2-D/1-D synthesis method. The MPACT code requires a cross section library to support all the MPACT core simulation capabilities which would be the most influencing component for simulation accuracy.}, author={Kim, Kang Seog ; and Clarno, Kevin T. ; and Gentry, Cole ; and Wiarda, Dorothea ; and Williams, Mark L. ; and Kochunas, Brendan ; and Liu, Yuxuan ; and Palmtag, Scott ; and Godfrey, Andrew T.}, year={2017}, month={Mar} }
 @book{kochunas_collins_stimpson_salko_jabaay_graham_liu_kim_wieselquist_godfrey_et al._2017, title={VERA Core Simulator methodology for pressurized water reactor cycle depletion}, url={https://www.osti.gov/scitech/biblio/1344991-vera-core-simulator-methodology-pressurized-water-reactor-cycle-depletion}, DOI={10.13182/NSE16-39}, abstractNote={This paper describes the methodology developed and implemented in the Virtual Environment for Reactor Applications Core Simulator (VERA-CS) to perform high-fidelity, pressurized water reactor (PWR), multicycle, core physics calculations. Depletion of the core with pin-resolved power and nuclide detail is a significant advance in the state of the art for reactor analysis, providing the level of detail necessary to address the problems of the U.S. Department of Energy Nuclear Reactor Simulation Hub, the Consortium for Advanced Simulation of Light Water Reactors (CASL). VERA-CS has three main components: the neutronics solver MPACT, the thermal-hydraulic (T-H) solver COBRA-TF (CTF), and the nuclide transmutation solver ORIGEN. This paper focuses on MPACT and provides an overview of the resonance self-shielding methods, macroscopic-cross-section calculation, two-dimensional/one-dimensional (2-D/1-D) transport, nuclide depletion, T-H feedback, and other supporting methods representing a minimal set of the capabilities needed to simulate high-fidelity models of a commercial nuclear reactor. Results are presented from the simulation of a model of the first cycle of Watts Bar Unit 1. The simulation is within 16 parts per million boron (ppmB) reactivity for all state points compared to cycle measurements, with an average reactivity bias of <5 ppmB for the entire cycle. Comparisons to cycle 1 flux map data are also provided, and the average 2-D root-mean-square (rms) error during cycle 1 is 1.07%. To demonstrate the multicycle capability, a state point at beginning of cycle (BOC) 2 was also simulated and compared to plant data. The comparison of the cycle 2 BOC state has a reactivity difference of +3 ppmB from measurement, and the 2-D rms of the comparison in the flux maps is 1.77%. These results provide confidence in VERA-CS's capability to perform high-fidelity calculations for practical PWR reactor problems.}, journal={Nuclear Science and Engineering}, author={Kochunas, Brendan ; and Collins, Benjamin ; and Stimpson, Shane ; and Salko, Robert ; and Jabaay, Daniel ; and Graham, Aaron ; and Liu, Yuxuan ; and Kim, Kang Seog ; and Wieselquist, William ; and Godfrey, Andrew ; and et al.}, year={2017}, month={Jan} }
 @book{graham_downar_palmtag_2016, title={Assessment of Thermal Hydraulic Feedback Models}, url={https://www.osti.gov/scitech/biblio/1325432-assessment-thermal-hydraulic-feedback-models}, author={Graham, Aaron ; and Downar, Thomas ; and Palmtag, Scott}, year={2016}, month={Jan} }
 @book{salko_blyth_dances_magedanz_jernigan_kelly_toptan_gergar_gosdin_avramova_et al._2016, title={CTF Validation and Verification Manual}, url={https://www.osti.gov/scitech/biblio/1342678-ctf-validation-verification-manual}, DOI={10.2172/1342678}, abstractNote={Coolant-Boiling in Rod Arrays- Two Fluids (COBRA-TF) is a Thermal/Hydraulic (T/H) simulation code designed for Light Water Reactor (LWR) analysis. It uses a two-fluid, three-field (i.e. fluid film, fluid drops, and vapor) modeling approach. Both sub-channel and 3D Cartesian forms of nine conservation equations are available for LWR modeling. The code was originally developed by Pacific Northwest Laboratory in 1980 and has been used and modified by several institutions over the last several decades. COBRA-TF is also used at the Pennsylvania State University (PSU) by the Reactor Dynamics and Fuel Management Group (RDFMG), and has been improved, updated, and subsequently became the PSU RDFMG version of COBRA-TF (CTF). One part of the improvement process includes validating the methods in CTF. This document seeks to provide a certain level of certainty and confidence in the predictive capabilities of the code for the scenarios it was designed to model--rod bundle geometries with operating conditions that are representative of prototypical Pressurized Water Reactor (PWR)s and Boiling Water Reactor (BWR)s in both normal and accident conditions. This is done by modeling a variety of experiments that simulate these scenarios and then presenting a qualitative and quantitative analysis of the results that demonstrates the accuracy to which CTF is capable of capturing specific quantities of interest.}, author={Salko, Robert K. ; and Blyth, Taylor S. ; and Dances, Christopher A. ; and Magedanz, Jeffrey W. ; and Jernigan, Caleb ; and Kelly, Joeseph ; and Toptan, Aysenur ; and Gergar, Marcus ; and Gosdin, Chris ; and Avramova, Maria ; and et al.}, year={2016}, month={May} }
 @book{collins_hamilton_jarrett_kim_kochunas_liu_palmtag_salko_stimpson_toth_et al._2016, title={Demonstrate VERA Core Simulator Performance Improvements L2:PHI.P13.03}, url={https://www.osti.gov/scitech/biblio/1338536-demonstrate-vera-core-simulator-performance-improvements-l2-phi-p13}, DOI={10.2172/1338536}, abstractNote={This report describes the performance improvements made to the VERA Core Simulator (VERA-CS) during FY2016. The development of the VERA Core Simulator has focused on the capability needed to deplete physical reactors and help solve various problems; this capability required the accurate simulation of many operating cycles of a nuclear power plant. The first section of this report introduces two test problems used to assess the run-time performance of VERA-CS using a source dated February 2016. The next section provides a brief overview of the major modifications made to decrease the computational cost. Following the descriptions of the major improvements, the run-time for each improvement is shown. Conclusions on the work are presented, and further follow-on performance improvements are suggested.}, author={Collins, Benjamin S. ; and Hamilton, Steven P. ; and Jarrett, Michael G. ; and Kim, Kang Seog ; and Kochunas, Brendan ; and Liu, Yuxuan ; and Palmtag, Scott ; and Salko, Robert K. ; and Stimpson, Shane G. ; and Toth, Alex ; and et al.}, year={2016}, month={Aug} }
 @book{collins_downar_fitzgerald_gehin_godfrey_graham_jabaay_kelley_kim_kochunas_et al._2016, title={MPACT Standard Input User s Manual, Version 2.2.0}, url={https://www.osti.gov/scitech/biblio/1342674-mpact-standard-input-user-manual-version}, DOI={10.2172/1342674}, abstractNote={The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.}, author={Collins, Benjamin S. ; and Downar, Thomas ; and Fitzgerald, Andrew ; and Gehin, Jess C. ; and Godfrey, Andrew T. ; and Graham, Aaron ; and Jabaay, Daniel ; and Kelley, Blake W. ; and Kim, Kang ; and Kochunas, Brendan ; and et al.}, year={2016}, month={Aug} }
 @book{downar_collins_gehin_godfrey_jabaay_kelley_clarno_kim_kochunas_larsen_et al._2016, title={MPACT Theory Manual, Version 2.2.0}, url={https://www.osti.gov/scitech/biblio/1340449-mpact-theory-manual-version}, DOI={10.2172/1340449}, abstractNote={This theory manual describes the three-dimensional (3-D) whole-core, pin-resolved transport calculation methodology employed in the MPACT code. To provide sub-pin level power distributions with sufficient accuracy, MPACT employs the method of characteristics (MOC) solutions in the framework of a 3-D coarse mesh finite difference (CMFD) formulation. MPACT provides a 3D MOC solution, but also a 2D/1D solution in which the 2D planar solution is provided by MOC and the axial coupling is resolved by one-dimensional (1-D) lower order (diffusion or P3) solutions. In Chapter 2 of the manual, the MOC methodology is described for calculating the regional angular and scalar fluxes from the Boltzmann transport equation. In Chapter 3, the 2D/1D methodology is described, together with the description of the CMFD iteration process involving dynamic homogenization and solution of the multigroup CMFD linear system. A description of the MPACT depletion algorithm is given in Chapter 4, followed by a discussion of the subgroup and ESSM resonance processing methods in Chapter 5. The final Chapter 6 describes a simplified thermal hydraulics model in MPACT.}, author={Downar, Thomas ; and Collins, Benjamin S. ; and Gehin, Jess C. ; and Godfrey, Andrew T. ; and Jabaay, Daniel ; and Kelley, Blake W. ; and Clarno, Kevin T. ; and Kim, Kang ; and Kochunas, Brendan ; and Larsen, Edward W. ; and et al.}, year={2016}, month={Jun} }
 @book{collins_downar_fitzgerald_gehin_godfrey_graham_jabaay_kelley_kim_kochunas_et al._2016, title={MPACT VERA Input User s Manual, Version 2.2.0}, url={https://www.osti.gov/scitech/biblio/1342675-mpact-vera-input-user-manual-version}, DOI={10.2172/1342675}, abstractNote={The MPACT (Michigan PArallel Charactistics based Transport) code is designed to perform high-fidelity light water reactor (LWR) analysis using whole-core pin-resolved neutron transport calculations on modern parallel-computing hardware. The code consists of several libraries which provide the functionality necessary to solve steady-state eigenvalue problems. Several transport capabilities are available within MPACT including both 2-D and 3-D Method of Characteristics (MOC). A three-dimensional whole core solution based on the 2D-1D solution method provides the capability for full core depletion calculations.}, author={Collins, Benjamin S. ; and Downar, Thomas ; and Fitzgerald, Andrew ; and Gehin, Jess C. ; and Godfrey, Andrew T. ; and Graham, Aaron ; and Jabaay, Daniel ; and Kelley, Blake W. ; and Kim, Kang ; and Kochunas, Brendan ; and et al.}, year={2016}, month={Jun} }
 @book{clarno_palmtag_secker_kendrick_montgomery_2016, title={Simulation of CRUD Induced Power Shift using the VERA Core Simulator and MAMBA}, url={https://www.osti.gov/scitech/biblio/1325433-simulation-crud-induced-power-shift-using-vera-core-simulator-mamba}, author={Clarno, Kevin T ; and Palmtag, Scott ; and Secker, Jeffrey ; and Kendrick, Brian ; and Montgomery, Rosemary}, year={2016}, month={Jan} }
 @book{godfrey_collins_kim_montgomery_powers_salko_stimpson_wieselquist_clarno_gehin_et al._2016, title={VERA Benchmarking Results for Watts Bar Nuclear Plant Unit 1 Cycles 1-12}, url={https://www.osti.gov/scitech/biblio/1260079-vera-benchmarking-results-watts-bar-nuclear-plant-unit-cycles}, author={Godfrey, Andrew T ; and Collins, Benjamin S ; and Kim, Kang Seog ; and Montgomery, Rosemary ; and Powers, Jeffrey J ; and Salko, Robert K ; and Stimpson, Shane G ; and Wieselquist, William A ; and Clarno, Kevin T ; and Gehin, Jess C ; and et al.}, year={2016}, month={Jan} }
 @book{salko_lange_kucukboyaci_sung_palmtag_gehin_avramova_2015, title={Development of COBRA-TF for Modeling Full-Core Reactor Operating Cycles}, url={https://www.osti.gov/scitech/biblio/1185922-development-cobra-tf-modeling-full-core-reactor-operating-cycles}, author={Salko, Robert K ; and Lange, Travis L ; and Kucukboyaci, Vefa ; and Sung, Yixing ; and Palmtag, Scott ; and Gehin, Jess C ; and Avramova, Maria}, year={2015}, month={Jan} }
 @book{salko_palmtag_collins_kendrick_seker_2015, title={L3.PHI.CTF.P10.02-rev2 Coupling of Subchannel T/H (CTF) and CRUD Chemistry (MAMBA1D)}, url={https://www.osti.gov/scitech/biblio/1344982-l3-phi-ctf-p10-rev2-coupling-subchannel-ctf-crud-chemistry-mamba1d}, DOI={10.2172/1344982}, abstractNote={The purpose of this milestone is to create a preliminary capability for modeling light water reactor (LWR) thermal-hydraulic (T/H) and CRUD growth using the CTF subchannel code and the subgrid version of the MAMBA CRUD chemistry code, MAMBA1D. In part, this is a follow-on to Milestone L3.PHI.VCS.P9.01, which is documented in Report CASL-U-2014-0188-000, titled "Development of CTF Capability for Modeling Reactor Operating Cycles with Crud Growth". As the title suggests, the previous milestone set up a framework for modeling reactor operation cycles with CTF. The framework also facilitated coupling to a CRUD chemistry capability for modeling CRUD growth throughout the reactor operating cycle. To demonstrate the capability, a simple CRUD \surrogate" tool was developed and coupled to CTF; however, it was noted that CRUD growth predictions by the surrogate were not considered realistic. This milestone builds on L3.PHI.VCS.P9.01 by replacing this simple surrogate tool with the more advanced MAMBA1D CRUD chemistry code. Completing this task involves addressing unresolved tasks from Milestone L3.PHI.VCS.P9.01, setting up an interface to MAMBA1D, and extracting new T/H information from CTF that was not previously required in the simple surrogate tool. Speci c challenges encountered during this milestone include (1) treatment of the CRUD erosion model, which requires local turbulent kinetic energy (TKE) (a value that CTF does not calculate) and (2) treatment of the MAMBA1D CRUD chimney boiling model in the CTF rod heat transfer solution. To demonstrate this new T/H, CRUD modeling capability, two sets of simulations were performed: (1) an 18 month cycle simulation of a quarter symmetry model of Watts Bar and (2) a simulation of Assemblies G69 and G70 from Seabrook Cycle 5. The Watts Bar simulation is merely a demonstration of the capability. The simulation of the Seabrook cycle, which had experienced CRUD-related fuel rod failures, had actual CRUD-scrape data to compare with results. As results show, the initial CTF/MAMBA1D-predicted CRUD thicknesses were about half of their expected values, so further investigation will be required for this simulation.}, author={Salko, Robert K. ; and Palmtag, Scott ; and Collins, Benjamin S. ; and Kendrick, Brian ; and Seker, Jeffrey}, year={2015}, month={May} }
 @book{kochunas_collins_jabaay_kim_graham_stimpson_wieselquist_clarno_palmtag_downar_et al._2015, title={VERA Core Simulator Methodology for PWR Cycle Depletion}, url={https://www.osti.gov/scitech/biblio/1215576-vera-core-simulator-methodology-pwr-cycle-depletion}, author={Kochunas, Brendan ; and Collins, Benjamin S ; and Jabaay, Daniel ; and Kim, Kang Seog ; and Graham, Aaron ; and Stimpson, Shane ; and Wieselquist, William A ; and Clarno, Kevin T ; and Palmtag, Scott ; and Downar, Thomas ; and et al.}, year={2015}, month={Jan} }
 @book{kouchunas_jabaay_downar_collins_stimpson_godfrey_kim_gehin_palmtag_franceschini_2015, title={Validation and Application of the 3D Neutron Transport MPACT within CASL VERA-CS}, url={https://www.osti.gov/scitech/biblio/1214019-validation-application-neutron-transport-mpact-within-casl-vera-cs}, author={Kouchunas, Brendan ; and Jabaay, Dan ; and Downar, Thomas ; and Collins, Benjamin S ; and Stimpson, Shane G ; and Godfrey, Andrew T ; and Kim, Kang Seog ; and Gehin, Jess C ; and Palmtag, Scott ; and Franceschini, Fausto}, year={2015}, month={Jan} }
 @book{schmidt_belcourt_hooper_pawlowski_clarno_simunovic_slattery_turner_palmtag_2014, title={An approach for coupled-code multiphysics core simulations from a common input}, url={https://www.osti.gov/scitech/biblio/1265624-approach-coupled-code-multiphysics-core-simulations-from-common-input}, DOI={10.1016/j.anucene.2014.11.015}, abstractNote={This paper describes an approach for coupled-code multiphysics reactor core simulations that is being developed by the Virtual Environment for Reactor Applications (VERA) project in the Consortium for Advanced Simulation of Light-Water Reactors (CASL). In this approach a user creates a single problem description, called the “VERAIn” common input file, to define and setup the desired coupled-code reactor core simulation. A preprocessing step accepts the VERAIn file and generates a set of fully consistent input files for the different physics codes being coupled. The problem is then solved using a single-executable coupled-code simulation tool applicable to the problem, which is built using VERA infrastructure software tools and the set of physics codes required for the problem of interest. The approach is demonstrated by performing an eigenvalue and power distribution calculation of a typical three-dimensional 17 × 17 assembly with thermal–hydraulic and fuel temperature feedback. All neutronics aspects of the problem (cross-section calculation, neutron transport, power release) are solved using the Insilico code suite and are fully coupled to a thermal–hydraulic analysis calculated by the Cobra-TF (CTF) code. The single-executable coupled-code (Insilico-CTF) simulation tool is created using several VERA tools, including LIME (Lightweight Integrating Multiphysics Environment for coupling codes), DTK (Data Transfer Kit), Trilinos, and TriBITS. Parallel calculations are performed on the Titan supercomputer at Oak Ridge National Laboratory using 1156 cores, and a synopsis of the solution results and code performance is presented. Ongoing development of this approach is also briefly described.}, journal={Annals of Nuclear Energy (Oxford)}, author={Schmidt, Rodney ; and Belcourt, Kenneth ; and Hooper, Russell ; and Pawlowski, Roger P. ; and Clarno, Kevin T. ; and Simunovic, Srdjan ; and Slattery, Stuart R. ; and Turner, John A. ; and Palmtag, Scott}, year={2014}, month={Dec} }
 @book{clarno_palmtag_davidson_salko_evans_turner_belcourt_hooper_schmidt_2014, title={Coupled Neutronics Thermal-Hydraulic Solution of a Full-Core PWR Using VERA-CS}, url={https://www.osti.gov/scitech/biblio/1154814-coupled-neutronics-thermal-hydraulic-solution-full-core-pwr-using-vera-cs}, author={Clarno, Kevin T ; and Palmtag, Scott ; and Davidson, Gregory G ; and Salko, Robert K ; and Evans, Thomas M ; and Turner, John A ; and Belcourt, Kenneth ; and Hooper, Russell ; and Schmidt, Rodney}, year={2014}, month={Jan} }
 @inproceedings{salko_avramova_hooper_palmtag_popov_turner_2013, title={Improvements, enhancements, and optimizations of COBRA-TF}, volume={3}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-84883365179&partnerID=MN8TOARS}, booktitle={International Conference on Mathematics and Computational Methods Applied to Nuclear Science and Engineering, M and C 2013}, author={Salko, R.K. and Avramova, M.N. and Hooper, R. and Palmtag, S. and Popov, E. and Turner, J.}, year={2013}, pages={2170–2181} }
 @inproceedings{mertyurek_palmtag_gert_finch_2010, title={Validation of the lattice physics code LANCER02 with ENDF/B-VII library}, volume={2}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-79952378953&partnerID=MN8TOARS}, booktitle={International Conference on the Physics of Reactors 2010, PHYSOR 2010}, author={Mertyurek, U. and Palmtag, S. and Gert, G. and Finch, J.}, year={2010}, pages={1130–1139} }
 @inproceedings{palmtag_lamas_finch_godfrey_moore_2008, title={The advanced BWR core simulator AETNA02}, volume={2}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-79953837410&partnerID=MN8TOARS}, booktitle={International Conference on the Physics of Reactors 2008, PHYSOR 08}, author={Palmtag, S. and Lamas, J. and Finch, J. and Godfrey, A. and Moore, B.R.}, year={2008}, pages={1282–1289} }
 @inproceedings{palmtag_mertyurek_moore_sugawara_toishigawa_2007, title={Steady-state nuclear analysis methods at global nuclear fuel, invited}, volume={97}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-55349115707&partnerID=MN8TOARS}, booktitle={Transactions of the American Nuclear Society}, author={Palmtag, S. and Mertyurek, U. and Moore, B.R. and Sugawara, M. and Toishigawa, A.}, year={2007}, pages={557–558} }
 @inproceedings{casmo-4 and multigroup mcnp comparisons for mox fuel assemblies_2005, volume={92}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-33244479401&partnerID=MN8TOARS}, booktitle={Transactions of the American Nuclear Society}, year={2005}, pages={508–509} }
 @inproceedings{bahadir_lindahl_palmtag_2005, title={Microscopic depletion model in SIMULATE-4}, volume={92}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-33244493781&partnerID=MN8TOARS}, booktitle={Transactions of the American Nuclear Society}, author={Bahadir, T. and Lindahl, S.-Ö. and Palmtag, S.P.}, year={2005}, pages={635–636} }
 @book{palmtag_1995, title={A two-dimensional, semi-analytic expansion method for nodal calculations}, url={https://www.osti.gov/scitech/biblio/505709-two-dimensional-semi-analytic-expansion-method-nodal-calculations}, DOI={10.2172/505709}, abstractNote={Most modern nodal methods used today are based upon the transverse integration procedure in which the multi-dimensional flux shape is integrated over the transverse directions in order to produce a set of coupled one-dimensional flux shapes. The one-dimensional flux shapes are then solved either analytically or by representing the flux shape by a finite polynomial expansion. While these methods have been verified for most light-water reactor applications, they have been found to have difficulty predicting the large thermal flux gradients near the interfaces of highly-enriched MOX fuel assemblies. A new method is presented here in which the neutron flux is represented by a non-seperable, two-dimensional, semi-analytic flux expansion. The main features of this method are (1) the leakage terms from the node are modeled explicitly and therefore, the transverse integration procedure is not used, (2) the corner point flux values for each node are directly edited from the solution method, and a corner-point interpolation is not needed in the flux reconstruction, (3) the thermal flux expansion contains hyperbolic terms representing analytic solutions to the thermal flux diffusion equation, and (4) the thermal flux expansion contains a thermal to fast flux ratio term which reduces the number of polynomial expansion functions needed to represent the thermal flux. This new nodal method has been incorporated into the computer code COLOR2G and has been used to solve a two-dimensional, two-group colorset problem containing uranium and highly-enriched MOX fuel assemblies. The results from this calculation are compared to the results found using a code based on the traditional transverse integration procedure.}, author={Palmtag, Scott P.}, year={1995}, month={Aug} }
 @book{wagner_redmond_palmtag_hendricks_1994, title={MCNP: Multigroup/adjoint capabilities}, volume={4}, url={https://www.osti.gov/scitech/biblio/10145615-mcnp-multigroup-adjoint-capabilities}, DOI={10.2172/10145615}, abstractNote={This report discusses various aspects related to the use and validity of the general purpose Monte Carlo code MCNP for multigroup/adjoint calculations. The increased desire to perform comparisons between Monte Carlo and deterministic codes, along with the ever-present desire to increase the efficiency of large MCNP calculations has produced a greater user demand for the multigroup/adjoint capabilities. To more fully utilize these capabilities, we review the applications of the Monte Carlo multigroup/adjoint method, describe how to generate multigroup cross sections for MCNP with the auxiliary CRSRD code, describe how to use the multigroup/adjoint capability in MCNP, and provide examples and results indicating the effectiveness and validity of the MCNP multigroup/adjoint treatment. This information should assist users in taking advantage of the MCNP multigroup/adjoint capabilities.}, institution={Office of Scientific and Technical Information (OSTI)}, author={Wagner, J.C.; and Redmond, E.L. II; and Palmtag, S.P.; and Hendricks, J.S.}, year={1994}, month={Apr} }
 @book{trybus_collins_maddison_bunde_palmtag_palmiotti_1994, title={Measurements of actinide transmutation in the hard spectrum of a fast reactor}, url={https://www.osti.gov/scitech/biblio/204716-measurements-actinide-transmutation-hard-spectrum-fast-reactor}, author={Trybus, C L; and Collins, P J; and Maddison, D W; and Bunde, K A; and Palmtag, S ; and Palmiotti, G}, year={1994}, month={Nov} }