@article{vaglio-gaudard_destouches_hawari_avramova_ivanov_valentine_blaise_hudelot_2023, title={Challenge for the validation of high-fidelity multi-physics LWR modeling and simulation: Development of new experiments in research reactors}, volume={11}, ISSN={["2296-598X"]}, DOI={10.3389/fenrg.2023.1110979}, abstractNote={Current approaches to validate multi-physics coupling mainly rely upon experimental data from the operation of the current reactor fleet. These data allow global experimental validation based on Light Water Reactor (LWR) macroscopic physical parameters of interest. However, they are insufficient for validating detailed coupling at the assembly and pin level. The use of well-controlled experimental data provided by research reactors is essential to implement a rigorous and consistent step-wise validation process of high-fidelity multi-physics coupling. That is why experimental data, such as the core power evolution in a transient-state coming from the SPERT-III experimental program and the CABRI research reactor, are analyzed as a first step towards this objective for the simulation of LWR transients initiated by reactivity insertion. The analysis of the state-of-the-art shows no existing experimental benchmark available worldwide for LWRs to consistently and rigorously validate advanced reactor physics/thermal-hydraulics/fuel performance coupling at the pin- or sub-channel scale. In this context, a discussion is therefore initiated in this paper on the perspective of developing new experiments dedicated to high-fidelity multi-physics tools, focusing on a first application: the validation of reactivity feedback effects. Very few existing light-water experimental reactors containing UO2fuel could today have the capacity to host these experiments. The development of a new validation experiment could only be achievable by considering a two-stage process for the experiment realization: a first stage involving a distributed network of sensors in the reactor core using instrumentation commonly used in research reactors, and a second stage implementing an instrumented fuel pin and innovative experimental techniques, in the longer term. Even if the OECD/NEA activities in the Expert Group on Multi-Physics Experimental Data, Benchmarks and Validation (EGMPEBV) (currently merged in the Expert Group on Multi-Physics of Reactor Systems – EGMUP) have started to pave the way for the development of such a high-fidelity multi-physics experiment, most of the work is still ahead of us.}, journal={FRONTIERS IN ENERGY RESEARCH}, author={Vaglio-Gaudard, Claire and Destouches, Christophe and Hawari, Ayman and Avramova, Maria and Ivanov, Kostadin and Valentine, Timothy and Blaise, Patrick and Hudelot, Jean-Pascal}, year={2023}, month={Jan} }
 @article{takasugi_aly_holler_abarca_beeler_avramova_ivanov_2023, title={Development of an efficient and improved core thermal-hydraulics predictive capability for fast reactors: Summary of research and development activities at the North Carolina state University}, volume={412}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2023.112474}, abstractNote={The improved understanding of the safety, technical gaps, and major uncertainties of advanced fast reactors will result in designing their safe and economical operation. This paper focuses on the development of efficient and improved core thermal-hydraulics predictive capabilities for fast reactor modeling and simulation at the North Carolina State University. The described research and development activities include applying results of high-fidelity thermal-hydraulic simulations to inform the improved use of lower-order models within fast-running design and safety analysis tools to predict improved estimates of local safety parameters for efficient evaluation of realistic safety margins for fast reactors. The above-described high-to-low model information improvements are being verified and validated using benchmarks such as the OECD/NRC Liquid Metal Fast Reactor Core Thermal-Hydraulic Benchmark and code-to-code comparisons.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Takasugi, C. and Aly, A. and Holler, D. and Abarca, A. and Beeler, B. and Avramova, M. and Ivanov, K.}, year={2023}, month={Oct} }
 @article{faure_delipei_petruzzi_avramova_ivanov_2023, title={Fuel performance uncertainty quantification and sensitivity analysis in the presence of epistemic and aleatoric sources of uncertainties}, volume={11}, ISSN={["2296-598X"]}, DOI={10.3389/fenrg.2023.1112978}, abstractNote={Fuel performance modeling and simulation includes many uncertain parameters from models to boundary conditions, manufacturing parameters and material properties. These parameters exhibit large uncertainties and can have an epistemic or aleatoric nature, something that renders fuel performance code-to-code and code-to-measurements comparisons for complex phenomena such as the pellet cladding mechanical interaction (PCMI) very challenging. Additionally, PCMI and other complex phenomena found in fuel performance modeling and simulation induce strong discontinuities and non-linearities that can render difficult to extract meaningful conclusions form uncertainty quantification (UQ) and sensitivity analysis (SA) studies. In this work, we develop and apply a consistent treatment of epistemic and aleatoric uncertainties for both UQ and SA in fuel performance calculations and use historical benchmark-quality measurement data to demonstrate it. More specifically, the developed methodology is applied to the OECD/NEA Multi-physics Pellet Cladding Mechanical Interaction Validation benchmark. A cold ramp test leading to PCMI is modeled. Two measured quantities of interest are considered: the cladding axial elongation during the irradiations and the cladding outer diameter after the cold ramp. The fuel performance code used to perform the simulation is FAST. The developed methodology involves various steps including a Morris screening to decrease the number of uncertain inputs, a nested loop approach for propagating the epistemic and aleatoric sources of uncertainties, and a global SA using Sobol indices. The obtained results indicate that the fuel and cladding thermal conductivities as well as the cladding outer diameter uncertainties are the three inputs having the largest impact on the measured quantities. More importantly, it was found that the epistemic uncertainties can have a significant impact on the measured quantities and can affect the outcome of the global sensitivity analysis.}, journal={FRONTIERS IN ENERGY RESEARCH}, author={Faure, Quentin and Delipei, Gregory and Petruzzi, Alessandro and Avramova, Maria and Ivanov, Kostadin}, year={2023}, month={Mar} }
 @article{takasugi_martin_laboure_ortensi_ivanov_avramova_2023, title={Preservation of kinetics parameters generated by Monte Carlo calculations in two-step deterministic calculations}, volume={9}, ISSN={["2491-9292"]}, DOI={10.1051/epjn/2022056}, abstractNote={The generation of accurate kinetic parameters such as mean generation time Λ and effective delayed neutron fraction βeff via Monte Carlo codes is established. Employing these in downstream deterministic codes warrants another step to ensure no additional error is introduced by the low-order transport operator when computing forward and adjoint fluxes for bilinear weighting of these parameters. Another complexity stems from applying superhomogenization (SPH) equivalence in non-fundamental mode approximations, where reference and low-order calculations rely on a 3D full core model. In these cases, SPH factors can optionally be computed for only part of the geometry while preserving reaction rates and K-effective, but the impact of such approximations on kinetics parameters has not been thoroughly studied. This paper aims at studying the preservation of bilinearly-weighted quantities in the Serpent–Griffin calculation procedure. Diffusion and transport evaluations of IPEN/MB-01, Godiva, and Flattop were carried out with the Griffin reactor physics code, testing available modeling options using Serpent-generated multigroup cross sections and equivalence data. Verifying Griffin against Serpent indicates sensitivities to multigroup energy grid selection and regional application of SPH equivalence, introducing significant errors; these were demonstrated to be reduced through the use of a transport method together with a finer energy grid.}, journal={EPJ NUCLEAR SCIENCES & TECHNOLOGIES}, author={Takasugi, Cole and Martin, Nicolas and Laboure, Vincent and Ortensi, Javier and Ivanov, Kostadin and Avramova, Maria}, year={2023}, month={Feb} }
 @article{aly_beeler_avramova_2022, title={Ab initio molecular dynamics investigation of gamma-(U,Zr) structural and thermal properties as a function of temperature and composition}, volume={561}, ISSN={["1873-4820"]}, DOI={10.1016/j.jnucmat.2022.153523}, abstractNote={Uranium in its metallic form is considered as a fuel for sodium fast reactors due to its higher thermal conductivity and high fissile material density relative to UO 2 fuel. The metal is alloyed with zirconium to increase its stability at high temperatures and increase its solidus temperature . This work uses ab initio molecular dynamics to perform an evaluation of the mechanical and thermophysical properties of the γ -(U,Zr) system at temperatures between 1000 K and 1400 K. Among these properties are the equilibrium volume, bulk modulus , molar heat capacity , heat of formation , and the surface energy. The obtained results are compared to experimental data and previous computational work available in the literature. This is the first study of γ -(U,Zr) utilizing ab initio molecular dynamics, and reduces thermophysical property knowledge gaps that are currently present in the literature.}, journal={JOURNAL OF NUCLEAR MATERIALS}, author={Aly, Ahmed and Beeler, Benjamin and Avramova, Maria}, year={2022}, month={Apr} }
 @article{delipei_rouxelin_abarca_hou_avramova_ivanov_2022, title={CTF-PARCS Core Multi-Physics Computational Framework for Efficient LWR Steady-State, Depletion and Transient Uncertainty Quantification}, volume={15}, ISSN={["1996-1073"]}, url={https://doi.org/10.3390/en15145226}, DOI={10.3390/en15145226}, abstractNote={Best Estimate Plus Uncertainty (BEPU) approaches for nuclear reactor applications have been extensively developed in recent years. The challenge for BEPU approaches is to achieve multi-physics modeling with an acceptable computational cost while preserving a reasonable fidelity of the physics modeled. In this work, we present the core multi-physics computational framework developed for the efficient computation of uncertainties in Light Water Reactor (LWR) simulations. The subchannel thermal-hydraulic code CTF and the nodal expansion neutronic code PARCS are coupled for the multi-physics modeling (CTF-PARCS). The computational framework is discussed in detail from the Polaris lattice calculations up to the CTF-PARCS coupling approaches. Sampler is used to perturb the multi-group microscopic cross-sections, fission yields and manufacturing parameters, while Dakota is used to sample the CTF input parameters and the boundary conditions. Python scripts were developed to automatize and modularize both pre- and post-processing. The current state of the framework allows the consistent perturbation of inputs across neutronics and thermal-hydraulics modeling. Improvements to the standard thermal-hydraulics modeling for such coupling approaches have been implemented in CTF to allow the usage of 3D burnup distribution, calculation of the radial power and the burnup profile, and the usage of Santamarina effective Doppler temperature. The uncertainty quantification approach allows the treatment of both scalar and functional quantities and can estimate correlation between the multi-physics outputs of interest and up to the originally perturbed microscopic cross-sections and yields. The computational framework is applied to three exercises of the LWR Uncertainty Analysis in Modeling Phase III benchmark. The exercises cover steady-state, depletion and transient calculations. The results show that the maximum fuel centerline temperature across all exercises is 2474K with 1.7% uncertainty and that the most correlated inputs are the 238U inelastic and elastic cross-sections above 1 MeV.}, number={14}, journal={ENERGIES}, author={Delipei, Gregory K. and Rouxelin, Pascal and Abarca, Agustin and Hou, Jason and Avramova, Maria and Ivanov, Kostadin}, year={2022}, month={Jul} }
 @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{lin_athe_rouxelin_avramova_gupta_youngblood_lane_dinh_2022, title={Digital-twin-based improvements to diagnosis, prognosis, strategy assessment, and discrepancy checking in a nearly autonomous management and control system}, volume={166}, ISSN={["1873-2100"]}, url={http://www.scopus.com/inward/record.url?eid=2-s2.0-85115958204&partnerID=MN8TOARS}, DOI={10.1016/j.anucene.2021.108715}, abstractNote={The Nearly Autonomous Management and Control System (NAMAC) is a comprehensive control system that assists plant operations by furnishing control recommendations to operators in a broad class of situations. This study refines a NAMAC system for making reasonable recommendations during complex loss-of-flow scenarios with a validated Experimental Breeder Reactor II simulator, digital twins improved by machine-learning algorithms, a multi-attribute decision-making scheme, and a discrepancy checker for identifying unexpected recommendation effects. We assess the performance of each NAMAC component, while we demonstrate and evaluated the capability of NAMAC in a class of loss-of-flow scenarios.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Lin, Linyu and Athe, Paridhi and Rouxelin, Pascal and Avramova, Maria and Gupta, Abhinav and Youngblood, Robert and Lane, Jeffrey and Dinh, Nam}, year={2022}, month={Feb} }
 @article{altahhan_geemert_avramova_ivanov_2022, title={Extending a low-order inhomogeneous adjoint equations model to a higher-order model with verification on integral applications}, volume={177}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2022.109277}, abstractNote={• Development and verification of a NEM-M2B2 mathematical inhomogeneous-adjoint nodal diffusion solver. • Application of the Lagrangian multipliers method to derive the nodal mathematical inhomogeneous-adjoint. • Derived the local linear prediction formula, specific for the forward NEM-M2B2 model, and utilized it to study the repercussions of perturbations in the IAEA-3D benchmark on the Axial Offset (AO). • Verified the generalized adjoint code developed while showing detailed steps of how inhomogeneous adjoint codes can be verified. • Compared between the low-order inhomogeneous NEM-M0 adjoint and the developed higher order inhomogeneous NEM-M2B2 model for the AO as a RoI. • Introduced the Mantissa theory to explain the behavior of the linear adjoint models and prediction formulas. A higher-order nodal mathematical inhomogeneous adjoint model conjugate to the NEM-M2B2 nodal diffusion forward model is developed and introduced in this research. Verification of the developed model is presented through applications in perturbation analysis and the IAEA-3D benchmark including adjusted forms of it. This paper’s objective is to explore ways of extending and optimizing a mathematical adjoint capability suitable for use in an industrial reactor code, such that it becomes not merely an approximate but rather the exact adjoint counterpart to the typically used higher-order nodal forward solvers used in mature industrial reactor codes. Specifically, it is investigated how to upgrade an already available lower-order nodal mathematical adjoint solver towards higher-order accuracy. An example of the latter is the lower-order nodal adjoint solver used in the ARTEMIS reactor code, in the technical context of stabilization and acceleration of embedded control rod search mechanisms. Though the latter adjoint solver proved suitable for the needed preconditioning purposes, while also enabling the benefit of computationally very lean adjoint iterations, several future developments could benefit from having a higher-order adjoint nodal solver available as well. By using a preconditioned form of the base NEM-M2B2 nodal diffusion forward model and by using variational analysis, we have obtained a higher-order nodal mathematical adjoint that can have a physical interpretation associated with it as a Lagrangian multiplier. The nodal mathematical adjoint is then developed for the Axial Offset (AO) as a Response of Interest (RoI) which leads to an inhomogeneous adjoint system of equations. A solution verification of the adjoint developed is done through analyzing the effects coming from perturbations in the absorption and the scattering cross-sections. The applications investigated include axially and radially traveling perturbations along the reactor’s core. Several locations for the traveling perturbations are chosen to represent important locations in the core. Comparison between the low-order and the higher-order adjoint models is conducted. The forward model is set to the NEM-M2B2 nodal diffusion equation for both adjoints during the comparison. The higher-order adjoint model developed show consistent results in comparison to its lower-order sibling, suggesting the preference of using the developed higher-order model for adjoint computations.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Altahhan, Muhammad Ramzy and Geemert, Rene and Avramova, Maria and Ivanov, Kostadin}, year={2022}, month={Nov} }
 @article{aly_beeler_avramova_2022, title={Investigation of ?-(U, Zr) structural properties and its interfacial properties with liquid sodium using ab initio molecular dynamics}, volume={567}, ISSN={["1873-4820"]}, url={https://doi.org/10.1016/j.jnucmat.2022.153835}, DOI={10.1016/j.jnucmat.2022.153835}, abstractNote={In this study, the elastic properties, structural parameters, sound velocity, and Debye temperature of γ−(U,Zr) were computed using ab initio molecular dynamics (AIMD) at temperatures between 1000 K and 1400 K and for Zr content between 0 at.% and 100 at.%. UZr is used as a metallic fuel for Sodium Fast Reactors (SFRs). The study of the mechanical and thermal behavior of these alloys leads to a better data-informed fuel design. The bulk modulus, shear modulus, Young’s modulus, and Poisson’s ratio were calculated from the elastic constants and their dependence on Zr content and temperature was investigated, comparing the results with previous computational work and the available experimental data in the literature. Interfacial properties between UZr (up to 32 at.% which typically exists in nuclear fuel) and liquid sodium are also of interest due to the presence of a sodium bond between the fuel and the cladding in metallic nuclear fuel. The interfacial energy between γ−(U,Zr) and liquid sodium, the surface tension of liquid sodium, and the work of adhesion were computed at different temperatures and Zr concentrations. It was demonstrated that γ−(U,Zr) is completely wetted by liquid sodium at all the investigated temperatures and Zr concentrations. This work provides the basis for the determination of interfacial resistances in SFRs and their implementation into heat transfer fuel performance simulations, which will be the subject of future work.}, journal={JOURNAL OF NUCLEAR MATERIALS}, publisher={Elsevier BV}, author={Aly, Ahmed and Beeler, Benjamin and Avramova, Maria}, year={2022}, month={Aug} }
 @article{casamor_avramova_reventos_freixa_2022, title={Off-line vs. semi-implicit TH-TH coupling schemes: A BEPU comparison}, volume={178}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2022.109344}, abstractNote={Several TH-TH code coupling methods are used in the nuclear industry to model the core thermal–hydraulic conditions and the system behaviour. In some cases, the boundary conditions obtained by system codes are applied to sub-channel codes by table (off-line coupling). Even though this approach is in general considered valid, some boundary parameters will present inconsistencies. Alternatively, system and sub-channel codes are coupled using different coupling methods (semi-implicit coupling). Recent studies have shown a strong influence of the boundary conditions uncertainty on the sub-channel code results. The present study aims to evaluate the differences produced by the coupling methods by performing a best-estimate plus uncertainties (BEPU) comparison to the following cases: a complete loss of forced flow and a pressurizer relief valve opening. Results show that BEPU analysis presents good agreement with some discrepancies that can be explained and correlated to the boundary conditions deviations between codes.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Casamor, M. and Avramova, M. and Reventos, F. and Freixa, J.}, year={2022}, month={Dec} }
 @article{rouxelin_alfonsi_strydom_avramova_ivanov_2022, title={Propagation of VHTRC manufacturing uncertainties with RAVEN/PHISICS}, volume={165}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2021.108667}, abstractNote={The International Atomic Energy Agency recently concluded a Coordinated Research Program (CRP) to evaluate the effect of propagation of uncertainties on design and safety parameters in High Temperature Gas-cooled Reactors (HTGRs). This CRP catalyzed the development of novel software and methods relevant to HTGR uncertainty analysis. In the framework of this CRP, the statistical analysis code RAVEN was coupled to the neutron transport code PHISICS, using 6-group cross section libraries generated with the modules TRITON/NEWT from SCALE 6.2.1. This article describes the mechanics of the RAVEN/PHISICS sequence, and reports the effects of manufacturing uncertainties on integral parameter uncertainties found in the Very High Temperature Reactor Critical (VHTRC) core. The VHTRC experimental results included propagation of manufacturing uncertainties to obtain eigenvalue (keff) and temperature coefficient (αT) uncertainties. RAVEN/PHISICS was used to reproduce this analysis and to compare the predicted output uncertainties to the experimental measurements on the three VHTRC cores (HC-I, HP, HC-II). Results from the sequence agree with the experimental values (σ[keff] ~ 0.00300). The analysis also focuses on the interpretation of input uncertainties. The simulations conducted with RAVEN/PHISICS demonstrated the input uncertainties can induce a threefold increase in the resulting output uncertainties, depending on the mathematical modeling of the raw input uncertainties. In particular, the use of a unique uncertainty value repeated over lattice elements constitutes the major contribution to the keff and αT uncertainties, while modeling these uncertainties with random independent values leads to negligible keff and αT uncertainties, due to cancellation of errors. The propagation of the manufacturing uncertainties was also repeated using 56 energy groups in the neutron transport calculations, and showed a moderate impact on the output (keff, αT) uncertainties (~10 % difference) compared to the base-case 6-group simulations.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Rouxelin, Pascal and Alfonsi, Andrea and Strydom, Gerhard and Avramova, Maria and Ivanov, Kostadin}, year={2022}, month={Jan} }
 @article{altahhan_geemert_avramova_ivanov_2021, title={Development and verification of a higher-order mathematical adjoint nodal diffusion solver}, volume={163}, ISSN={["1873-2100"]}, DOI={10.1016/j.anucene.2021.108548}, abstractNote={In this paper, we derive a mathematical formulation of the higher order adjoint NEM-M2B2 equations by preconditioning the nodal interface neutron currents equations of the forward equations system, and by using the Lagrangian Multipliers analysis method. In the NEM-M2B2 system of equations, the quadratic transverse leakage approximation is used to model the leakage of neutrons between each node in the system. The solution of the adjoint equation can be used to perform adjoint-based predictive sensitivity/perturbation analysis. As an example, we use the mathematical adjoint solution as sensitivity weighting for predicting the response of the IAEA-3D benchmark’s eigenvalue to a perturbation in the independent parameters of the system (i.e., cross-sections). We also derive perturbation equations associated with the particular NEM-M2B2 model we are using. These perturbation-equations are used in predicting the model eigenvalue change without resorting to recalculating the forward NEM-M2B2 system of equations again (labeled as exact calculations). They also enabled construction of a reactivity sensitivity map showing the importance of each calculation node of the benchmark depending on its spatial and spectral coordinates. Perturbations were imposed on both the absorption cross-sections (fast and thermal) and the scattering cross-section of the IAEA-3D benchmark problem. Several verification steps were taken to ensure that the developed mathematical adjoint solver is adequate for adjoint analysis (e.g., commutativity checks, and comparison against exact calculations).}, journal={ANNALS OF NUCLEAR ENERGY}, author={Altahhan, Muhammad Ramzy and Geemert, Rene and Avramova, Maria and Ivanov, Kostadin}, year={2021}, month={Dec} }
 @article{delipei_hou_avramova_rouxelin_ivanov_2021, title={Summary of comparative analysis and conclusions from OECD/NEA LWR-UAM benchmark Phase I}, volume={384}, ISSN={["1872-759X"]}, url={http://dx.doi.org/10.1016/j.nucengdes.2021.111474}, DOI={10.1016/j.nucengdes.2021.111474}, abstractNote={In recent years, large efforts have been devoted to Light Water Reactor (LWR) Uncertainty Quantification (UQ). In 2006, the LWR Uncertainty Analysis in Modeling (UAM) benchmark was launched with an aim to investigate the uncertainty propagation in all modeling stages of the LWRs and guide uncertainty and sensitivity analysis methodology development. This article summarizes the benchmark activities for the standalone neutronics phase (Phase I), which includes three main exercises: Exercise I-1: “Cell Physics,” Exercise I-2: “Lattice Physics,” and Exercise I-3: “Core Physics.” A comparative analysis of the Phase I results is performed in this article for all the considered LWRs types: Three Mile Island – 1 Pressurized Water Reactor (PWR), Peach Bottom – 2 Boiling Water Reactor (BWR), Kozloduy – 6 Water - Water Energetic Reactor (VVER) and a Generation-III reactor. It was found, for all major exercises, that the predicted uncertainty of the system eigenvalue is highly dependent on the choice of the covariance libraries used in the UQ process and is less sensitive to the solution method, nuclear data library and UQ method. For all four reactor types, the observed relative standard deviation across all exercises is approximately 0.5% for the UO2 fuel. In the pin cell and lattice calculations with MOX fuel this uncertainty increases to 1%. The main reason is the larger Pu-239 nu-bar uncertainty compared to the U-235 nu-bar. The largest contributors to the eigenvalue uncertainties are the U-235 nu-bar and the U-238 capture in the UO2 fuel and the Pu-239 nu-bar in the MOX fuel. In the assembly lattice exercises, higher uncertainties are predicted for the fast group than the thermal group constants with differences up to one order of magnitude. This is attributed to the larger uncertainties of most cross-sections at high energies. The obtained correlation matrices share some common major trends but also exhibit strong differences in case by case comparisons indicating an impact of the selected neutronics modeling and nuclear data library. In the core exercises, the predicted relative standard deviation of the radial and axial power, for most of the cores, is below 10%. An exception is the radial power profile of the Generation-III core, when a mixture of UOX/MOX assemblies is considered. Finally, it is important to note that the bias in most of the studies is significant and up to the same order of the estimated uncertainty. This indicates a need for better quantification of the bias/variance through more code to code and code to experiments comparisons.}, journal={NUCLEAR ENGINEERING AND DESIGN}, publisher={Elsevier BV}, author={Delipei, Gregory Kyriakos and Hou, Jason and Avramova, Maria and Rouxelin, Pascal and Ivanov, Kostadin}, year={2021}, month={Dec} }
 @article{avramova_2020, title={Developments in thermal-hydraulic sub-channel modeling for whole core multi-physics simulations}, volume={358}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.110387}, abstractNote={The focus of this article is on current developments of thermal-hydraulic sub-channel codes and associated nuclear fuel rod models for high-fidelity light water reactor (LWR) core multi-physics applications. The advanced thermal-hydraulics sub-channel code CTF is selected as an example. CTF is a shortened name given to a version of the legacy code COBRA-TF. Reactor Dynamics and Fuel Modeling Group at North Carolina State University and Oak Ridge National Laboratory are jointly developing CTF. The code has been integrated in the Consortium for Advanced Simulation of LWRs Virtual Environment for Reactor Applications code system. Within the CTF user’s group, the code is being used for different applications in academia, research institutions and industry. Recent developments and applications of CTF for LWR multi-physics modeling and simulation and associated verification, validation, and uncertainty quantification analyses are summarized. In addition, acceleration techniques and parallelization strategies for improving the computational efficiency for whole LWR core sub-channel calculations are briefly introduced. Special attention is given to the high-to-low fidelity model information approach for CTF improvements. Recent developments related to residual-based sub-channel equations and physics-informed data-driven modeling are discussed.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Avramova, Maria}, year={2020}, month={Mar} }
 @article{balestra_henry_carlyon_english_myer_avramova_epiney_strydom_2020, title={Modular high temperature gas reactor core modeling with RELAP5-3D/ PHISICS ? Optimization schemes for load following}, volume={362}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2020.110526}, abstractNote={The objective of this study is to develop a 3D neutron kinetic (NK) and thermal hydraulic (TH) coupled model using the INL developed code RELAP5-3D©/PHISICS to study the load following operation of a Modular High-Temperature Gas-Cooled Reactor (MHTGR). The selected design is the 350 MW prismatic, graphite moderated, helium cooled thermal reactor based on the MHTGR-350 transient benchmark led by the High-Temperature Gas-Cooled Reactor (HTGR) Methods Core Simulation Group at Idaho National Laboratory (INL) in the framework of the Organization for Economic Cooperation and Development Nuclear Energy Agency (OECD/NEA). This kind of reactor usually reacts very slowly to the perturbation of the core parameters due to the large amount of graphite in the core. This kind of behavior promotes the reactor stability but on the other hand limits the reactor load-following operability. Recently advances in gas reactor designs have made load-following a feasible and achievable goal. Modern nuclear reactors (such as the MHTGR-350) are designed to withstand the structural stresses associated with load-following. Operating the reactor in load-following mode will inevitably reduce the load factor. Although a higher load factor means more revenue and the best usage of the fuel, following the variable energy demand will increase the maximum achievable fraction of generated nuclear power, being no more limited to the base load power generation. In general, if low impact on material aging and the safe operability are demonstrated, the plant economics will be minimally affected. In order to ensure that the system can be safely operated in a load-following mode, an extensively study has been carried out. Some tests with linear change in coolant mass flow rate demonstrated that the reactor behavior is suitable for operation in load following mode. This assumption has been confirmed by a 4 days load following transient test in which the reactor supplied the requested power with a negligible error.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Balestra, Paolo and Henry, Keion and Carlyon, Cameron and English, Cody and Myer, Jaylon and Avramova, Maria and Epiney, Aaron and Strydom, Gerhard}, year={2020}, month={Jun} }
 @article{toptan_salko_avramova_clarno_kropaczek_2019, title={A new fuel modeling capability, CTFFuel, with a case study on the fuel thermal conductivity degradation}, volume={341}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2018.11.010}, abstractNote={A new fuel modeling capability, CTFFuel, is developed from the subchannel code, CTF. This code is a standalone interface to the CTF fuel rod models, allowing for fuel rod simulations to be run independently from the fluid. This paper provides an overview of the code with a case study on the thermal conductivity degradation of LWR fuels to demonstrate its capabilities. The modeling of fuel thermal conductivity degradation in the code is improved through the addition of new modeling options to account for the irradiation effects via globally defined parameters. After the initial implementation, a variety of order-of-accuracy tests and code comparisons are performed to test software quality. A controlled analysis is allowed by CTFFuel to verify the numerical scheme of CTF’s conduction solution and to benchmark its fuel temperature predictions against FRAPCON-4.0’s. Overall, the software quality and verification procedure ensures that the new model is coded correctly, that it properly interacts with the rest of the code.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Salko, Robert K. and Avramova, Maria N. and Clarno, Kevin and Kropaczek, David J.}, year={2019}, month={Jan}, pages={248–258} }
 @article{bennett_martin_avramova_2019, title={A surrogate model based on sparse grid interpolation for boiling water reactor subchannel void distribution}, volume={131}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2019.03.022}, abstractNote={Boiling water reactors are simulated using nodal diffusion core simulators which rely upon homogenized and condensed cross sections from a lattice physics code. In the lattice calculation, the void distribution is typically assumed to be uniform in the radial direction. To remove this assumption, a thermal hydraulic code can be coupled with a lattice physics code to include a radial void distribution in the cross sections. To minimize the additional computational costs, a surrogate model can be generated for the thermal hydraulic code. In this research, a surrogate model is generated for the thermal hydraulic code F-COBRA-TF using sparse grid interpolation. The surrogate model is tested on how well it can reproduce the F-COBRA-TF void distribution for various conditions on the ATRIUM 10 assembly. The surrogate model is found to be effective at reproducing the F-COBRA-TF void distribution and reducing the computational costs from the order of minutes to about a second. A coupling is created between the surrogate model and the lattice physics code APOLLO2-A. Including the void distribution in the lattice physics calculation is found to have a large effect on the gadolinium worth.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Bennett, Alexander and Martin, Nicolas and Avramova, Maria}, year={2019}, month={Sep}, pages={51–62} }
 @article{porter_mousseau_avramova_2019, title={CTF-R: A novel residual-based thermal hydraulic solver}, volume={348}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.04.006}, abstractNote={The traditional scientific process has been revolutionized by the advent of computational modeling, but the nuclear industry generally uses “legacy codes,” which were developed early in the evolution of computers. One example of a legacy code, the thermal hydraulic subchannel code CTF, is modernized in this work through the development of a novel residual-based version, CTF-R. Unlike its predecessor, CTF-R is not limited by the strict computational limitations of the early 1980’s, and can therefore be designed such that it is inherently flexible and easy to use. A case study is examined to demonstrate how the flexibility of the code can be used to improve simulation results. In this example, the coupling between the solid and liquid fields is examined. Traditionally, this coupling is modeled explicitly, which imposes numerical stability limits on the time step size. These limits are derived and it is shown that they are removed when the coupling is made implicit. Further, the development of CTF-R will enable future improvements in next generation reactor modeling, numerical methods, and coupling to other codes. Through the further development of CTF-R and other residual-based codes, state-of-the-art simulation is possible.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Porter, N. W. and Mousseau, V. A. and Avramova, M. N.}, year={2019}, month={Jul}, pages={37–45} }
 @article{toptan_kropaczek_avramova_2019, title={Gap conductance modeling I: Theoretical considerations for single- and multi-component gases in curvilinear coordinates}, volume={353}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.110283}, abstractNote={Abstract Accurate estimation of heat transfer across the gap is important in nuclear fuel performance because heat transfer across the fuel-to-cladding gap heavily impacts fuel temperatures and the thermo-mechanical performance of nuclear fuel rods. Better understood physics will allow a better prediction of the gap behavior. This paper focuses on providing an overview of the gap conductance model including theoretical considerations and underlying assumptions. The gap conductance is calculated considering three summed heat paths: fill gas conductance, direct thermal radiation, and solid contact conductance. Each heat transfer mechanism is described in detail. First, the models are generalized to curvilinear coordinates for diatomic/polyatomic molecules. Traditional models use one-dimensional Cartesian equations for a monatomic gas. Second, expressions for temperature jump distance and thermal accommodation coefficients are made consistent with the kinetic theory for both single- and multi-component gases. Lastly, fill gas thermal conductivity is updated to include its dependence on rod internal pressure.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Kropaczek, David J. and Avramova, Maria N.}, year={2019}, month={Nov} }
 @article{toptan_kropaczek_avramova_2019, title={Gap conductance modeling II: Optimized model for UO2-Zircaloy interfaces}, volume={355}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.110289}, abstractNote={The model conventionally used to calculate heat transfer across the fuel-cladding gap in light water nuclear reactors is a modified version of the Ross-Stoute model. The model was modified to include gap distance in the formulation, which introduced additional uncertainties because the model parameters were not adjusted after the modification. In this study, this conventional model is optimized for uranium dioxide-Zircaloy interfaces using experimental data at high pressure for single- and multi-component gases. First, a calibration is performed for single-component gases. Second, the calibration is extended to multi-component gases, which allows for a demonstration of sources of uncertainty in the model. Third, a general form of the gap conductance model is optimized by combining both data sets. Difficulties arise due to: (i) inaccurate estimation of contact characteristics (e.g., number of solid contacts, deformation mechanism of surface irregularities, contact shapes) that are different for each experimental setup; (ii) the non-physical ratio of temperature jump distance to the gap distance for postulated model function form; (iii) an insufficient description of the appropriate heat transfer regime; and (iv) the pressure dependence of thermal conductivity for inert gases aside from helium. Lastly, a general model is optimized by setting the temperature jump distance at the wall to zero, which reduces possible uncertainties. This final analysis results in a more accurate prediction of the available experimental data. The Associated parameter uncertainty of the model is estimated by performing uncertainty propagation. Overall, the optimized model results in a larger gap conductance with significantly reduced error.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Kropaczek, David J. and Avramova, Maria N.}, year={2019}, month={Dec} }
 @article{toptan_kropaczek_avramova_2019, title={On the validity of the dilute gas assumption for gap conductance calculations in nuclear fuel performance codes}, volume={350}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2019.04.042}, abstractNote={Fill gas thermal conductivity’s dependence on pressure is neglected in today’s nuclear fuel performance codes. Current codes assume that gas behaves as a dilute gas, but the pressure effect is more pronounced at temperatures lower than ten times the critical temperature of each pure gas. The validity of this assumption for nuclear fuel performance is examined herein. Theories related to dilute and dense gas properties are presented, along with their validation against literature data at up to 30 MPa for selected inert gases. Underlying assumptions are clearly described for each model, and their possible impacts on gap conductance calculations are discussed. The dilute gas assumption is valid for helium because it behaves as a dilute gas. However, the assumption is not valid in most gap conductance calculations when the gap is mostly occupied with either lower conductivity gaseous fission products or an initial fill gas other than helium.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Toptan, Aysenur and Kropaczek, David J. and Avramova, Maria N.}, year={2019}, month={Aug}, pages={1–8} }
 @article{porter_mousseau_avramova_2019, title={Quantified Validation with Uncertainty Analysis for Turbulent Single-Phase Friction Models}, volume={205}, ISSN={["1943-7471"]}, DOI={10.1080/00295450.2018.1548221}, abstractNote={Abstract This paper introduces a framework for model selection that includes parameter estimation, uncertainty propagation, and quantified validation. The framework is applied to single-phase turbulent friction modeling in CTF, which is a thermal-hydraulic code for nuclear engineering applications. The friction model is chosen because it is well understood and easy to separate from other physics, which allows focus to be on the model selection framework instead of on the particulars of the chosen model. Two different empirical models are compared: the McAdams Correlation and the Simplified McAdams Correlation. The parameter estimation is performed by calibrating each of the friction models to experimental data using the Delayed Rejection Adaptive Metropolis algorithm, which is a Markov Chain Monte Carlo method. State point uncertainties are also considered, which are determined based on measurement errors from the experiment. The input parameter distributions are propagated through CTF using a statistical method with samples. A variety of validation metrics is used to quantify which empirical model is more accurate. It is shown that model form uncertainty can be quantified using validation once all other sources of uncertainty—numerical, sampling, experimental, and parameter—have been quantitatively addressed. When multiple models are available, the one that has the smallest model form error can be selected. Though the framework is applied to a simple example here, the same process can quantify the model form uncertainty of more complicated physics, multiple models, and simulation tools in other fields. Therefore, this work is a demonstration of best practices for future assessments of model form uncertainty.}, number={12}, journal={NUCLEAR TECHNOLOGY}, author={Porter, Nathan W. and Mousseau, Vincent A. and Avramova, Maria N.}, year={2019}, month={Dec}, pages={1607–1617} }
 @article{gosdin_avramova_2018, title={Application of sub-channel modeling to BWR core analysis}, volume={115}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2018.02.005}, abstractNote={The current trends in nuclear science and engineering, related to modeling and simulation, are towards high-fidelity multi-physics multi-scale simulations to address industry challenge and high impact problems. These trends stimulate the utilization of sub-channel modeling approaches to nuclear reactor cores for both operation applications (core follow and cycle depletion evaluations) and safety applications (transient and accident analysis). While sub-channel modeling of Pressurized Water Reactor (PWR) cores has advanced significantly in last few years, the application of sub-channel modeling to Boiling Water Reactor (BWR) cores is under development. CTF is an improved version of COBRA-TF being developed and maintained by the North Carolina State University (NCSU) in cooperation with Oak Ridge National Laboratory (ORNL), and with support of the US Department of Energy (DOE) Consortium for Advanced Simulation of Light Water Reactors (CASL) as well as from the members of the CTF User’s Group. CTF uses a two-fluid, three-field representation of the two-phase flow, which makes it capable of modeling the high-void flow conditions expected in BWR operation. This paper focuses on applications of CTF to mini- and whole-core BWR calculations on assembly/channel and pin-cell/sub-channel resolved levels as well as on demonstrating that CTF can properly model bypass flow. To increase the confidence in CTF’s BWR modeling capabilities, simulations have been performed using the international Organization for Economic Cooperation and Development (OECD)/US Nuclear Regulatory Commission (NRC) Oskarshamn-2 benchmark, including modeling of a single assembly and a mini-core of 2 × 2 assemblies on a pin-by-pin/sub-channel level, and a full core model on an assembly/channel level. Each model is varied with an increasing amount of detail. Key parameters such as pressure losses and void fraction distribution were analyzed to determine the impact of different levels of detail within a thermal-hydraulic model on the simulation results. The results demonstrated that CTF is capable of modeling BWR core on different spatial resolution levels. The Oskarshamn-2 core simulations was used to further verify and demonstrate CTF’s capabilities of modeling BWRs.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Gosdin, C. and Avramova, M.}, year={2018}, month={May}, pages={294–302} }
 @article{li_avramova_jiao_chen_yu_pu_chen_2018, title={CFD prediction of critical heat flux in vertical heated tubes with uniform and non-uniform heat flux}, volume={326}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2017.11.009}, abstractNote={In this paper, Eulerian Two-fluid model coupled with extended wall boiling model was used to simulate the departure from nucleate boiling (DNB) in vertical heated tubes under high pressures by using STAR-CCM+ 10.04. Based on CFD approach, new methods were developed to predict the critical heat flux (CHF) in both uniform and non-uniform heated tubes. The results showed some differences between the uniform and non-uniform heated cases. The transition in boiling curves was used as the criterion of DNB for uniform heat flux while the peaks of wall temperature and near wall void fraction were used for non-uniform cases. Good agreement was obtained between the predictions and experimental data, including both the values of critical heat flux and their locations. For uniform heat flux, sensitivity analysis for pressure, local quality and mass flow rate were performed, and the results showed that the new prediction method could work well under a wide range of conditions except under the high local quality and low mass flow rate conditions. The mechanism of this prediction method was discussed as well.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Li, Quan and Avramova, M. and Jiao, Yongjun and Chen, Ping and Yu, Junchong and Pu, Zengping and Chen, Jie}, year={2018}, month={Jan}, pages={403–412} }
 @article{toptan_porter_salko_avramova_2018, title={Implementation and assessment of wall friction models for LWR core analysis}, volume={115}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2018.02.022}, abstractNote={The modeling of frictional pressure drop in the nuclear thermal hydraulics subchannel code, CTF, is improved through the addition of three new modeling options. Two of the new models allow the code to account for the effects of surface roughness and the third enables a user-supplied option. After the initial implementation, a variety of analyses are performed to test the software quality. First, a series of defect tests are designed for both single- and multi-channel configurations which compare simulated results to approximate solutions. The single-channel tests assess the friction model implementation; a suite of three-by-three bundle tests are used to ensure proper implementation of the roughness averaging scheme. The maximum relative error in the pressure drop over all defect tests is less than 0.15%. A solution verification test is performed to ensure that the first order numerical scheme in CTF is not significantly disrupted by the friction model. Finally, the wall friction model is validated using both separate and integral effects experimental data. Overall, the software quality, verification, and validation procedure ensures that the new model is coded correctly, that it properly interacts with the rest of CTF, and that it can be used to model real-world data for turbulent single-phase flow. The work completed herein provides a complete demonstration of modern coding practices. Future work could include a formal equation analysis of the numerical error in the friction model, as well as an analysis of validation data for one dimensional two-phase flow.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Toptan, Aysenur and Porter, Nathan W. and Salko, Robert K. and Avramova, Maria N.}, year={2018}, month={May}, pages={565–572} }
 @article{mankosa_avramova_2018, title={Three-dimensional multi-physics modeling of hydrogen and hydride distribution in zirconium alloy cladding}, volume={105}, ISSN={["0149-1970"]}, DOI={10.1016/j.pnucene.2018.02.012}, abstractNote={Localized phenomena within the reactor core, specifically those associated with the nuclear fuel, require high-fidelity simulations to enable accurate physics predictions. One example is the zirconium cladding, which absorbs hydrogen from the light water coolant during normal reactor operation. Absorbed in the cladding, this hydrogen is in solid solution and its distribution is sensitive to temperature and concentration gradients. At high enough concentrations, the hydrogen will precipitate as a hydride. Thus, the hydrogen distribution as a hydride precipitate in cladding has been identified as an important safety concern, and a possible ersatz for validating reactor simulation code temperature models. This study reports development efforts of using high-fidelity multi-physics codes to model temperature, hydrogen, and hydride distributions in three dimensions under realistic operating conditions. The Consortium for the Advanced Simulation of Light Water Reactors multi-physics code, Tiamat, is used to model selected sub-assemblies. Then, a single fuel pin is selected from the sub-assembly and modeled as a three-dimensional BISON problem. The outer cladding temperatures from the Tiamat calculation are used as boundary conditions for the BISON problem in order to obtain hydrogen and hydride distributions. Areas of interest for hydride precipitation include locations along the fuel rod experiencing highest temperatures with significant spatial variation, particularly in the vicinity of the spacer grids and mixing vanes.}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Mankosa, Michael and Avramova, Maria}, year={2018}, month={May}, pages={294–300} }
 @article{porter_avramova_mousseau_2018, title={Uncertainty Quantification Study of CTF for the OECD/NEA LWR Uncertainty Analysis in Modeling Benchmark}, volume={190}, ISSN={["1943-748X"]}, DOI={10.1080/00295639.2018.1435135}, abstractNote={Abstract This work describes the results of a quantitative uncertainty analysis of the thermal-hydraulic subchannel code for nuclear engineering applications, Coolant Boiling in Rod Arrays-Three Field (COBRA-TF). CTF is used, which is a version of COBRA-TF developed in cooperation between the Consortium for Advanced Simulation of Light Water Reactors and North Carolina State University. Four steady-state cases from Phase II Exercise 3 of the Organisation for Economic Co-operation and Development/Nuclear Energy Agency Light Water Reactor Uncertainty Analysis in Modeling (UAM) Benchmark are analyzed using the statistical analysis tool, Design Analysis Kit for Optimization and Terascale Applications (Dakota). The input parameters include boundary condition, geometry, and modeling uncertainties, which are selected using a sensitivity study and then defined based on expert judgment. A forward uncertainty quantification method with Latin hypercube sampling (LHS) is used, where the sample size is based on available computational resources. The means and standard deviations of thermal-hydraulic quantities of interest are reported, as well as the Spearman rank correlation coefficients between the inputs and outputs. The means and standard deviations are accompanied by their respective standard errors, and the correlation coefficients are tested for statistical significance. The quantities of interest include void fractions, temperatures, and pressure drops. The predicted uncertainty in all parameters remains relatively low for all quantities of interest. The dominant sources of uncertainty are identified. For cases based on experiments, two different validation metrics are used to quantify the difference between measured and predicted void fractions. The results compare well with past studies, but with a number of improvements: the use of an updated CTF input deck using the current UAM specification and the most recent version of CTF, the use of an LHS method, an analysis of standard errors for the statistical results, and a quantitative comparison to experimental data. Though the statistical uncertainty analysis framework presented herein is applied to thermal-hydraulic analyses, it is generally applicable to any simulation tool. Given a specified amount of computational resources, it can be used to quantify statistical significance through the use of fundamental statistical analyses. This is in contrast with the prevailing methods in nuclear engineering, which provide a sample size necessary to achieve a specified level of statistical certainty.}, number={3}, journal={NUCLEAR SCIENCE AND ENGINEERING}, author={Porter, Nathan W. and Avramova, Maria N. and Mousseau, Vincent A.}, year={2018}, pages={271–286} }
 @article{porter_avramova_2018, title={Validation of CTF pressure drop and void predictions for the NUPEC BWR database}, volume={337}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2018.07.018}, abstractNote={A quantified validation of CTF pressure drop, equilibrium quality, and void fraction predictions is performed using the Japanese Nuclear Power Engineering Corporation boiling water reactor database. Four quantities of interest are compared between the experiments and the code predictions: pressure drop, average exit equilibrium quality, average exit void fraction, and subchannel exit void fraction. These four quantities of interest have root-mean-squared errors of 0.124, 0.005, 0.065, and 0.089, respectively. Pressure drop predictions are generally underpredicted for single phase cases and overpredicted for two phase cases. The equilibrium quality predictions are mostly within 0.01 of the designed experimental values, which indicates proper energy conservation. The void fraction results tend to be overpredicted by about 0.06, which is attributed to the interfacial modeling in the code. By splitting the subchannels into different groups, it is shown that those near unheated surfaces are the least accurately modeled, especially for cases that have a high concentration of unheated surfaces. The results are consistent with past validation analyses using a variety of subchannel codes. Unlike past studies with CTF, this work incorporates all 69 pressure drop experiments and 392 steady state void experiments from the database. A new method for quantification of asymmetries is proposed and applied, but the interpretation of the results depends on the definition of the measurement error.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Porter, Nathan W. and Avramova, Maria N.}, year={2018}, month={Oct}, pages={291–299} }
 @article{raja_mankosa_avramova_2017, title={Comparative analysis of different fidelity fuel performance models for fuel temperature predictions}, volume={322}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2017.07.025}, abstractNote={Thermal-hydraulics subchannel models have proven to give an acceptable compromise between modeling fidelity and computational efficiency in coupled multi-physics steady state, depletion, and transient calculations. The accuracy in modeling both gas-gap conductance and fuel thermal conductivity is of significant importance for fuel temperature prediction, which, in turn, is crucial for calculation of Doppler feedback on power. A comparative analysis of fuel performance models of different fidelity was performed using the sub-channel code CTF, and the higher fidelity fuel performance codes FRAPCON and BISON. The purpose of this study was to ascertain the predictive accuracy of the informed fuel rod model in CTF, thus the potential to inform this model using data from higher fidelity fuel performance models. Excellent agreement was found between CTF and FRAPCON, and between CTF and BISON with respect to inside clad temperature as well as to fuel surface and fuel centerline temperatures when the CTF gap conductance was set to the BISON and FRAPCON calculated gap conductance values. With respect to the CTF and FRAPCON comparison, the maximum temperature difference between the two codes for a given power level and burnup value was below 2 degree Kelvin for clad inner surface and fuel surface temperature. For fuel centerline temperature, the maximum temperature difference was found to be below 7 degree Kelvin at the highest power level and burnup value. Similarly, the CTF and BISON comparison resulted in maximum temperature differences less than 5 degree Kelvin for the fuel centerline temperature. These results demonstrate that, if the gap conductance, dimensions, and radial power distribution is correctly set in CTF, the CTF-predicted rod temperature distribution will match closely with higher fidelity tools and licensed industry level fuel performance codes for normal operating conditions.}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Raja, Faisal and Mankosa, Michael and Avramova, Maria}, year={2017}, month={Oct}, pages={464–473} }
 @article{raja_avramova_2017, title={Evaluation of modeling options for in-pellet power distribution and gap gas conductance for accurate fuel temperature predictions}, volume={100}, ISSN={["0149-1970"]}, DOI={10.1016/j.pnucene.2017.06.005}, abstractNote={Having the ability to predict fuel temperatures for efficient multi-physics steady state, depletion, and transient calculations with reasonable accuracy without the added burden of prohibitively expensive computation costs has been a major driving force in the nuclear industry. There are several parameters that have an immense impact on fuel surface and centerline temperatures. Sensitivity studies were performed to investigate the impact of gap gas conductance and internal pin power distribution on the fuel temperature predictions. As a result, areas of improvement in the CTF fuel performance model were identified by separating different effects, and analyzing the sensitivity of results to each model improvement. The performed studies demonstrated the importance of modeling internal pellet power distribution for accurate prediction of fuel centerline temperature. Furthermore, a new gap gas conductance modeling option that leverages the fuel performance code FRAPCON was implemented in the fuel rod model of CTF. Gap gas conductance data was pre-computed as a function of linear heat rate and fuel exposure, and was integrated into CTF as part of the new model. Using FRAPCON as a reference solution, the new FRAPCON-informed gap conductance model of CTF was found to calculate results within 2 degrees Kelvin of FRAPCON predictions with respect to fuel surface temperature. This study indicated the feasibility of developing an efficient framework for informing the low fidelity fuel rod models of thermal-hydraulic codes, in this case CTF, with more accurate pre-computed values by leveraging high fidelity fuel performance codes such as FRAPCON. CTF was able to utilize this tabulated data provided by the FRAPCON fuel performance code as well as to include the above mentioned improvements in each axial node of a given rod to provide a full three-dimensional representation.}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Raja, Faisal and Avramova, Maria}, year={2017}, month={Sep}, pages={135–145} }
 @article{davis_courty_avramova_motta_2017, title={High-fidelity multi-physics coupling for determination of hydride distribution in Zr-4 cladding}, volume={110}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2017.06.049}, abstractNote={Hydride production in Zircaloy cladding continues to be one of the main limiting factors for extending the life of nuclear fuel rods in the core. The production of hydrides in the cladding comes as a direct result of corrosion with water during normal operation. Furthermore, the distribution of hydrogen in the cladding depends strongly on the temperature and temperature gradients inside the cladding. In order to accurately predict these temperature gradients, a high-fidelity multi-physics coupling is needed. The Department of Energy (DOE) recognized this need and sponsored a project at the Pennsylvania State University (PSU) in cooperation with North Carolina State University (NCSU) under the Nuclear Energy University Programs (NEUP). The overreaching goal of this project is to couple thermal-hydraulics, neutronics, and fuel performance codes together to predict the distribution of hydrogen and hydrides in the cladding as a function of time and space. This goal is attained through a two-step approach. The first step combines accurate high-fidelity thermal-hydraulic models for heat transfer, reactor physics models for neutron flux, and thermal-mechanics models for fuel performance calculations to acquire detailed temperature and stress distributions in the fuel rod. The second step develops a semi-analytical model and experimentally tests the temperature and/or stress dependent hydrogen pick-up, diffusion, and precipitation in the cladding. This paper aims to show the capabilities of the high-fidelity coupling, their effect on the power and temperature predictions, and subsequent effect on the distribution of hydrogen in the cladding, specifically in the inter-pellet gap region.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Davis, Ian and Courty, Olivier and Avramova, Maria and Motta, Arthur}, year={2017}, month={Dec}, pages={475–485} }
 @article{yilmaz_avramova_andersen_2017, title={Multi-physics code system with improved feedback modeling}, volume={98}, ISSN={["0149-1970"]}, DOI={10.1016/j.pnucene.2017.03.007}, abstractNote={Fuel temperature (Doppler) feedback modeling in the coupled sub-channel thermal-hydraulic/time dependent neutron transport codes system CTF/TORT-TD was improved by accounting for the burnup dependence of the fuel thermal conductivity. TORT-TD is a three-dimensional (3D) time dependent neutron-kinetics code based on the discrete ordinates (SN) method. CTF is the Reactor Dynamics and Fuel Modeling Group (RDFMG) version of the sub-channel thermal-hydraulics code COBRA-TF (COlant Boiling in Rod Arrays – Two Fluid). A burnup-dependent fuel rod model, which takes into account the degradation of the fuel thermal conductivity at high burnups and the effects of burnable poisons, such as Gadolinium, was implemented in CTF. The model is applicable to UO2 (uranium dioxide) and MOX (mixed oxide) nuclear fuels – it includes the modified Nuclear Fuel Industries (NFI) model for UO2 fuels and the Duriez/Modified NFI model for MOX fuels. The in-pellet fuel temperature distributions predicted by CTF/TORT-TD were compared to reference CTF/TORT-TD/FRAPCON calculations, in which the fuel rods were modeled with the fuel performance code FRAPCON. These comparisons were carried out for a 4 × 4 pressurized water reactor (PWR) pin array at hot full power (HFP) steady state conditions. The CTF/TORT-TD fuel temperature predictions were consistent with the CTF/TORT-TD/FRAPCON results. This fact demonstrated that CTF with the new fuel thermal conductivity model can predict the temperature field within light water reactor (LWR) fuel rods as accurately as FRAPCON. Therefore, CTF/TORT-TD calculations can be carried out in fast scoping studies instead of the computationally expensive CTF/TORT-TD/FRAPCON calculations. The performed statistical analyses indicated an improved accuracy of fuel temperature calculations relative to the CTF/TORT-TD/FRAPCON reference numerical solution. Furthermore, better agreement between CTF/TORT-TD and CTF/TORT-TD/FRAPCON in calculated neutronic reactivity was found when fuel burnup effects were considered in CTF/TORT-TD. Therefore, the improved CTF/TORT-TD can be seen as a high fidelity multi-physics computational tool capable of providing accurate and efficient simulations for practical reactor core design and safety analysis.}, journal={PROGRESS IN NUCLEAR ENERGY}, author={Yilmaz, Mine O. and Avramova, Maria N. and Andersen, Jens G. M.}, year={2017}, month={Jul}, pages={94–108} }
 @article{bennett_avramova_ivanov_2016, title={Coupled MCNP6/CTF code: Development, testing, and application}, volume={96}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2016.05.008}, abstractNote={This paper presents the development and testing of a high fidelity Monte Carlo based multi-physics code. The coupling was done between the Monte Carlo neutronics code MCNP6 and the thermal-hydraulic sub-channel code CTF. The coupling for the MCNP6/CTF code was done internally at the pin level. On-The-Fly cross sections were used to decrease the complexity of the coupled code as well as to decrease the memory requirement. The relaxation acceleration technique was applied to the coupled code and was shown that it could satisfy much stricter convergence criterions. The technique can also guarantee convergence and be used as a tool to decrease the computational time. The coupled code was tested against two other coupled Monte Carlo/thermal-hydraulic sub-channel codes and the results were similar. The coupled code was also tested on a full assembly problem.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Bennett, A. and Avramova, M. and Ivanov, K.}, year={2016}, month={Oct}, pages={1–11} }
 @article{yilmaz_avramova_andersen_2016, title={Development, verification, and validation of a fuel thermal conductivity degradation model in CTF}, volume={97}, ISSN={["0306-4549"]}, DOI={10.1016/j.anucene.2016.07.020}, abstractNote={This paper discusses the implementation of a burnup dependent fuel thermal conductivity model within the Reactor Dynamics and Fuel Modeling Group (RDFMG) version of the subchannel thermal-hydraulics code COBRA-TF (CTF). The model takes into account the degradation of fuel thermal conductivity at high burnups and its dependence on the Gadolinium content for both UO2 (uranium dioxide) and MOX (mixed oxide) nuclear fuel rods. The modified Nuclear Fuel Industries (NFI) model for UO2 rods and the Duriez/Modified NFI model for MOX rods were incorporated in CTF. To validate the models, the fuel centerline temperatures predicted with CTF were compared to Halden reactor experimental data and to high fidelity FRAPCON-3.4 calculations. Halden test cases for UO2 fuel rods at the beginning of life (BOL), through lifetime with and without Gd2O3; and for MOX fuel rods were simulated with CTF. It was demonstrated that CTF with the new burnup dependent fuel thermal conductivity model predicts the fuel centerline temperature with less than a 5% error as compared to the Halden measurements. CTF calculations were performed for fifty-eight (58) data points. Statistical analyses of the dimensionless predicted-to-measured fuel centerline temperature ratios had confirmed the advantage of the new model – the mean value of the predicted-to-measured temperature ratios was increased from 0.8920 to 1.0082 and the standard deviation was decreased from 0.0693 to 0.0382.}, journal={ANNALS OF NUCLEAR ENERGY}, author={Yilmaz, Mine O. and Avramova, Maria N. and Andersen, Jens G. M.}, year={2016}, month={Nov}, pages={246–261} }
 @article{yilmaz_avramova_andersen_2016, title={Impact of fuel thermal conductivity degradation on Doppler feedback during rod ejection accident}, volume={307}, ISSN={["1872-759X"]}, DOI={10.1016/j.nucengdes.2016.07.021}, abstractNote={This paper discusses the importance of the fuel thermal conductivity degradation modeling for accurate predictions of the Doppler feedback during reactivity insertion transients. The impact of the fuel thermal conductivity degradation model, recently implemented in the coupled sub-channel thermal-hydraulic/time-dependent neutron transport code system CTF/TORT-TD, on Doppler feedback predictions during a control rod ejection accident was investigated. The rod ejection was simulated for a 4 × 4 pressurized water reactor pin array, extracted from the Purdue University MOX (mixed oxide) benchmark, starting at both hot zero power and hot full power conditions with the control rod being half-inserted before the ejection. The two scenarios were simulated with CTF/TORT-TD and the effect of the fuel thermal conductivity degradation on the Doppler feedback was analyzed. The results were compared with existing reference calculations performed with the coupled sub-channel thermal-hydraulic/time-dependent neutron transport/fuel performance code system CTF/TORT-TD/FRAPCON-FRAPTRAN. The power pulse, the time evolution of average fuel temperature, and the peak enthalpy rise during the transient were examined. It was confirmed that the impact of the fuel thermal conductivity degradation is more significant when the control rod is ejected at hot full power conditions. If the fuel conductivity degradation was not taken into account, less conservative CTF/TORT-TD predictions for the transient power response were obtained. For the selected 4 × 4 pin array, the coupled code calculated 13 MW higher power pulse when modeling degradation effects on fuel conductivity. The difference in the power response is due to the less negative prompt fuel temperature (Doppler) coefficient at elevated temperatures. Lower thermal conductivity will lead to higher fuel pellet temperatures, and subsequently to a less negative Doppler coefficient, which will result in a stronger power pulse. The maximum fuel enthalpy rise during the hot full power rod ejection accident was found to be 60 cal/g (251,208 J/kg).}, journal={NUCLEAR ENGINEERING AND DESIGN}, author={Yilmaz, Mine O. and Avramova, Maria N. and Andersen, Jens G. M.}, year={2016}, month={Oct}, pages={339–353} }
 @article{magedanz_avramova_perin_velkov_2015, title={High-fidelity multi-physics system TORT-TD/CTF/FRAPTRAN for light water reactor analysis}, volume={84}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2015.01.033}, DOI={10.1016/J.ANUCENE.2015.01.033}, abstractNote={A need exists in the nuclear industry for higher-fidelity tools for light water reactor (LWR) analysis, due to increasing core heterogeneity and higher burnup of fuels. In order to address this need, a high-fidelity multi-physics (HFMP) system has been developed at the Pennsylvania State University (PSU). It consists of three codes – CTF for thermal hydraulics, TORT-TD for neutron kinetics, and FRAPTRAN for fuel performance. FRAPCON, which is applied to long-term steady-state fuel performance, is left separate and not modified, but is relevant to the system because it generates the initial conditions used in FRAPTRAN. These codes have been combined into a system in which they are coupled by means of serial integration. FRAPTRAN is the latest addition to the system while the initial coupling of TORT-TD and CTF was verified in different applications. Recent efforts have been directed at the design of an object-oriented system of interfaces for the coupled codes, by which the main program may control them in terms of high-level functionality. Further modifications to the system include the ability to use coolant-centered rather than fuel-centered channels, and the ability for TORT-TD to use a time step size that differs from that of CTF. The obtained results verify this new coupling, as well as demonstrate the advantages of using a fuel-performance code for modeling fuel rod feedback.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Magedanz, J. and Avramova, M. and Perin, Y. and Velkov, A.K.}, year={2015}, month={Oct}, pages={234–243} }
 @article{avramova_ivanov_kozlowski_pasichnyk_zwermann_velkov_royer_yamaji_gulliford_2015, title={Multi-physics and multi-scale benchmarking and uncertainty quantification within OECD/NEA framework}, volume={84}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.12.014}, DOI={10.1016/J.ANUCENE.2014.12.014}, abstractNote={• Presentation of latest multi-physics multi-scale NEA/OECD benchmarks. • Utilization of high-quality experimental data for detailed comparative analysis. • Including uncertainty and sensitivity analysis of modeling predictions. • Uncertainty propagation in LWR multi-physics and multi-scale simulations. The development of multi-physics multi-scale coupled methodologies for Light Water Reactor (LWR) analysis requires comprehensive validation and verification procedures, which include well-established benchmarks developed in international cooperation. The Nuclear Energy Agency (NEA) of the Organization for Economic Co-operation and Development (OECD) has provided such framework, and over the years a number of LWR benchmarks have been developed and successfully conducted. The first set of NEA/OECD benchmarks that permits testing of the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions have been completed and documented. These benchmarks provided a validation basis for the new generation of coupled “best-estimate” codes. The above mentioned OECD/NEA LWR benchmark activities have also stimulated follow up developments and benchmarks to test these developments. The models utilized have been improved when moving from one benchmark to the next and this created a need to validate them using high-quality experimental data. Second set of the NEA/OECD benchmarks have been initiated by the Expert Group on Uncertainty Analysis in Modelling (EGUAM) at the Nuclear Science Committee (NSC), NEA/OECD to address the current trends in the development of LWR multi-physics and multi-scale modeling and simulation. These benchmarks include the following common features, which address some of the issues identified in the first set of OECD/NEA benchmarks: (a) utilization of high-quality experimental data; (b) refined local scale modeling in addition to global predictions; (c) more detailed comparisons and analysis; (d) including uncertainty and sensitivity analysis of modeling predictions. The paper presents each of these new benchmarks by providing description and discussion of comparative analysis of obtained results. Special attention is devoted to uncertainty propagation in LWR multi-physics and multi-scale simulations for design and safety evaluations.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Avramova, M. and Ivanov, K. and Kozlowski, T. and Pasichnyk, I. and Zwermann, W. and Velkov, K. and Royer, E. and Yamaji, A. and Gulliford, J.}, year={2015}, month={Oct}, pages={178–196} }
 @article{salko_schmidt_avramova_2015, title={Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications}, volume={84}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.11.005}, DOI={10.1016/J.ANUCENE.2014.11.005}, abstractNote={This paper describes major improvements to the computational infrastructure of the CTF subchannel code so that full-core, pincell-resolved (i.e., one computational subchannel per real bundle flow channel) simulations can now be performed in much shorter run-times, either in stand-alone mode or as part of coupled-code multi-physics calculations. These improvements support the goals of the Department Of Energy Consortium for Advanced Simulation of Light Water Reactors (CASL) Energy Innovation Hub to develop high fidelity multi-physics simulation tools for nuclear energy design and analysis. A set of serial code optimizations—including fixing computational inefficiencies, optimizing the numerical approach, and making smarter data storage choices—are first described and shown to reduce both execution time and memory usage by about a factor of ten. Next, a “single program multiple data” parallelization strategy targeting distributed memory “multiple instruction multiple data” platforms utilizing domain decomposition is presented. In this approach, data communication between processors is accomplished by inserting standard Message-Passing Interface (MPI) calls at strategic points in the code. The domain decomposition approach implemented assigns one MPI process to each fuel assembly, with each domain being represented by its own CTF input file. The creation of CTF input files, both for serial and parallel runs, is also fully automated through use of a pressurized water reactor (PWR) pre-processor utility that uses a greatly simplified set of user input compared with the traditional CTF input. To run CTF in parallel, two additional libraries are currently needed: MPI, for inter-processor message passing, and the Parallel Extensible Toolkit for Scientific Computation (PETSc), which is used to solve the global pressure matrix in parallel. Results presented include a set of testing and verification calculations and performance tests assessing parallel scaling characteristics up to a full-core, pincell-resolved model of a PWR core containing 193 17 × 17 assemblies under hot full-power conditions. This model, representative of Watts Bar Unit 1 and containing about 56,000 pins, was modeled with roughly 59,000 subchannels, leading to about 2.8 million thermal–hydraulic control volumes in total. Results demonstrate that CTF can now perform full-core analysis of a PWR (not previously possible owing to excessively long runtimes and memory requirements) on the order of 20 min. This new capability not only is useful to stand-alone CTF users, but also is being leveraged in support of coupled code multi-physics calculations being done in the CASL program.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Salko, Robert K. and Schmidt, Rodney C. and Avramova, Maria N.}, year={2015}, month={Oct}, pages={122–130} }
 @article{rosenkrantz_avramova_ivanov_prinsloo_botes_elsakhawy_2014, title={Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR}, volume={73}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/J.ANUCENE.2014.06.018}, DOI={10.1016/J.ANUCENE.2014.06.018}, abstractNote={Abstract The purpose of this study was to develop a coupled accurate multi-physics model of the SAFARI-1 Material Testing Reactor (MTR), a facility that is used for both research and the production of medical isotopes. The model was developed as part of the SAFARI-1 benchmarking project as a cooperative effort between the Pennsylvania State University (PSU) and the South African Nuclear Energy Corporation (Necsa). It was created using a multi-physics coupling of state of the art nuclear reactor simulation tools, consisting of a neutronics code and a thermal hydraulics code. The neutronics tool used was the PSU code NEM, and the results from this component were verified using the Necsa neutronics code OSCAR-4, which is utilized for SAFARI-1 core design and fuel management. On average, the multiplication factors of the neutronics models agreed to within 5 pcm and the radial assembly-averaged powers agreed to within 0.2%. The thermal hydraulics tool used was the PSU version of COBRA-TF (CTF) sub-channel code, and the results of this component were verified against another thermal hydraulics code, the RELAP5-3D system code, used at Necsa for thermal–hydraulics analysis of SAFARI-1. Although only assembly-averaged results from RELAP5-3D were available, they fell within the range of values for the corresponding assemblies in the comprehensive CTF solution. This comparison allows for the first time to perform a quantification of steady-state errors for a low-powered MTR with an advanced thermal–hydraulic code such as CTF on a per-channel basis as compared to simpler and coarser-mesh RELAP5-3D modeling. Additionally, a new cross section representation was used to ensure that the thermal hydraulic feedback effects on the core neutronics were captured as accurately as possible. This cross section representation was applied to SAFARI-1 core calculations for the first time in this work. Such implementation helps to quantify the effect of detailed modeling of thermal–hydraulics feedback effects on neutronics results in multi-physics simulations. The outcome of the study is the intended coupled neutronics/thermal–hydraulics model of the SAFARI-1 reactor.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Rosenkrantz, Adam and Avramova, Maria and Ivanov, Kostadin and Prinsloo, Rian and Botes, Danniëll and Elsakhawy, Khalid}, year={2014}, month={Nov}, pages={122–130} }
 @article{biery_avramova_2014, title={Investigations on coupled code PWR simulations using COBRA-TF with soluble boron tracking model}, volume={77}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/J.PNUCENE.2014.06.005}, DOI={10.1016/J.PNUCENE.2014.06.005}, abstractNote={For long term reactivity control over a nuclear reactor core fuel cycle, Pressurized Water Reactors (PWRs) make use of chemical shim in the form of soluble boron added to the coolant water. While soluble boron allows for even reactivity control and more uniform fuel burn-up, maintaining uniform distribution of the boron is important to prevent localized transients. Transients that are caused by a local disturbance in the concentration of boron are classified as boron dilution transients. While many studies have been performed to study these types of transients, the choice of existing codes available to simulate soluble boron transport have required tradeoffs to be made. Popularly used system codes can only simulate one-dimensional boron transport with comparatively simple physical models, which neglect important physical characteristics of boron transport in the fluid such as mixing due to cross flow between channels and turbulence effects. On the other extreme, Computational Fluid Dynamics (CFD) codes are capable of modeling boron transport with very high fidelity, but most CFD codes still require a large amount of computational resources to simulate a realistic physical model. Recent work at the Pennsylvania State University (PSU) has helped to fill this capability gap. The result is an improvement to the PSU's version of COBRA-TF (PSU CTF) by employing a newly developed boron tracking model. The resulting version of CTF is known as CTF-BTM (CTF-Boron Tracking Model). The implemented boron tracking model uses a Modified Godunov method to solve the boron transport field equation. Although the CTF boron tracking model was rigorously tested at the time it was developed, it has not yet been used in coupled thermal hydraulics and neutronics simulations, which is the aim of this study. The objective of this study is to continue the verification and qualification of the boron tracking model used in CTF-BTM. This is accomplished by first coupling CTF-BTM to the nodal diffusion-based neutronics code NEM. Part II and Part III of the OECD/NRC PWR MOX/UO2 Core Transient Benchmark are then used to validate the coupled code at Hot-Full Power (HFP) conditions and Hot-Zero Power (HZP) conditions using the 2-group homogenized cross section data supplied by the benchmark (generated using 47-group HELIOS 1.7). Close agreement to the benchmark solutions is achieved in both cases. This study culminates in the execution of a postulated post-Small Break Loss Of Coolant Accident (SBLOCA) boron dilution accident and an accompanying sensitivity study. While the scenario is highly idealized in terms of initiating events and assumptions, it provides an example of one of the expected future applications of coupled CTF and three-dimensional neutronics codes in LWR simulations with the introduction of boron tracking capability. In this culminating scenario, a series of simulations are executed where deborated condensate water slugs are inserted into the core. The slugs are formed in the steam generators by condensation following loss of coolant inventory sufficient to maintain natural circulation. First, for the natural circulation cases the most limiting core location where the slug was expected to enter was found. The limiting case was determined by the peak fuel enthalpy in the hottest fuel node, which was well below the enthalpy level expected to cause fuel damage. For the same limiting location of condensate slug entry, the case is simulated again at forced circulation conditions assuming actuation of the corresponding reactor coolant pump. As expected, the core response was more rapid than exhibited in the natural circulation cases with a larger power excursion and peak fuel enthalpy. At this fuel enthalpy level, extensive fuel damage is expected. The observed power excursions are also highly localized- indicating loose coupling between the region of interest and the rest of the core.}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Biery, M. and Avramova, M.}, year={2014}, month={Nov}, pages={72–83} }
 @article{ozdemir_avramova_sato_2014, title={Multi-dimensional boron transport modeling in subchannel approach: Part I. Model selection, implementation and verification of COBRA-TF boron tracking model}, volume={278}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/J.NUCENGDES.2013.02.031}, DOI={10.1016/J.NUCENGDES.2013.02.031}, abstractNote={The risk of reflux condensation especially during a Small Break Loss Of Coolant Accident (SB-LOCA) and the complications of tracking the boron concentration experimentally inside the primary coolant system have stimulated and subsequently have been a focus of many computational studies on boron tracking simulations in nuclear reactors. This paper presents the development and implementation of a multidimensional boron transport model with Modified Godunov Scheme within a thermal-hydraulic code based on a subchannel approach. The cross flow mechanism in multiple-subchannel rod bundle geometry as well as the heat transfer and lateral pressure drop effects are considered in the performed studies on simulations of deboration and boration cases. The Pennsylvania State University (PSU) version of the COBRA-TF (CTF) code was chosen for the implementation of three different boron tracking models: First Order Accurate Upwind Difference Scheme, Second Order Accurate Godunov Scheme, and Modified Godunov Scheme. Based on the performed nodalization sensitivity studies, the Modified Godunov Scheme approach with a physical diffusion term was determined to provide the best solution in terms of precision and accuracy. As a part of the verification and validation activities, a code-to-code comparison was carried out with the STAR-CD computational fluid dynamics (CFD) code and presented here. The objective of this study was two-fold: (1) to verify the accuracy of the newly developed CTF boron tracking model against CFD calculations; and (2) to investigate its numerical advantages as compared to other thermal-hydraulics codes.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Ozdemir, Ozkan Emre and Avramova, Maria N. and Sato, Kenya}, year={2014}, month={Oct}, pages={701–712} }
 @article{ozdemir_avramova_2014, title={Multi-dimensional boron transport modeling in subchannel approach: Part II. Validation of CTF boron tracking model and adding boron precipitation model}, volume={278}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/J.NUCENGDES.2014.08.003}, DOI={10.1016/J.NUCENGDES.2014.08.003}, abstractNote={The risk of small-break loss of coolant accident (SB-LOCA) and other reactivity initiated transients caused by boron dilution in the light water reactors (LWRs), and the complications of tracking the soluble boron concentration experimentally inside the primary coolant have stimulated the interest in computational studies for accurate boron tracking simulations in nuclear reactors. In Part I of this study, the development and implementation of a multi-dimensional boron transport model with modified Godunov scheme based on a subchannel approach within the COBRA-TF (CTF) thermal-hydraulic code was presented. The modified Godunov scheme approach with a physical diffusion term was determined to provide the most accurate and precise solution. Current paper extends these conclusions and presents the model validation studies against experimental data from the Rossendorf coolant mixing model (ROCOM) test facility. In addition, the importance of the two-phase flow characteristics in modeling boron transient are emphasized, especially during long-term cooling period after the loss of coolant accident (LOCA) condition in pressurized water reactors (PWRs). The CTF capabilities of boron transport modeling are further improved based on the three-field representation of the two-phase flow utilized in the code. The boron transport within entrained droplets is modeled, and a model for predicting the boron precipitation under transient conditions is developed and tested. It is aimed to extend the applicability of CTF to reactor transient simulations, and particularly to a large-break loss of coolant accident (LB-LOCA) analysis.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Ozdemir, Ozkan Emre and Avramova, Maria N.}, year={2014}, month={Oct}, pages={713–722} }
 @article{clifford_ivanov_avramova_2013, title={A multi-scale homogenization and reconstruction approach for solid material temperature calculations in prismatic high temperature reactor cores}, volume={256}, ISSN={0029-5493}, url={http://dx.doi.org/10.1016/j.nucengdes.2012.11.016}, DOI={10.1016/j.nucengdes.2012.11.016}, abstractNote={Traditional full-core heat transfer analysis of high temperature reactors uses effective coarse mesh parameters that are typically derived from a priori analysis and/or simplified analytical models that approximate the subscale temperature response. Consequently, different assumptions are made on each spatial scale, potentially yielding inconsistent solution methods with large associated uncertainties. In contrast, homogenized cross-sections used in full-core neutronics analysis are obtained using consistent homogenization techniques applied in conjunction with unit cell calculations. This approach has been proven both efficient and accurate. It is therefore surprising that formal homogenization techniques are rarely used in heat transfer analysis of nuclear reactors. In this work we take advantage of distinct unit cells that can be identified on each spatial scale in the MHTGR reactor core with the view to develop a consistent and accurate methodology for constructing hierarchical coarse mesh models for solid heat conduction in this reactor type. Three techniques have been used: formal multi-scale expansion homogenization is applied to obtain effective unit cell thermodynamic parameters; coarse mesh temperature discontinuities are defined to ensure continuity of the fine-scale temperatures at interfaces; and reduced order models for the time-dependent temperature response of the unit cells are obtained using proper orthogonal decomposition applied to detailed unit cell simulation results. The result is an efficient method, aimed toward unstructured CFD frameworks, that accurately captures coarse mesh temperatures with the capability of fully reconstructing the fine scale solution at any hierarchical level. The advantages of this method are illustrated for a small prismatic HTGR core using a cascaded solution approach. Starting at the finest scale, the TRISO coated particles, high resolution unit cell calculations are performed in a hierarchical fashion to build up a library of homogenized coarse mesh parameters and reduced order models, which are then used for the full-core heat conduction analysis. We demonstrate the accuracy and efficiency of the method by comparing results for a typical HTR power excursion transient against detailed reference solutions.}, journal={Nuclear Engineering and Design}, publisher={Elsevier BV}, author={Clifford, Ivor and Ivanov, Kostadin N. and Avramova, Maria N.}, year={2013}, month={Mar}, pages={1–13} }
 @article{espel_avramova_ivanov_misu_2013, title={New developments of the MCNP/CTF/NEM/NJOY code system – Monte Carlo based coupled code for high accuracy modeling}, volume={51}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2012.06.031}, DOI={10.1016/j.anucene.2012.06.031}, abstractNote={High accuracy code systems are necessary to model core environments with considerable geometry complexity and great material heterogeneity. These features are typical of current and innovative nuclear reactor core designs. Advanced methodologies and state-of-the art coupled code systems must be put into practice in order to model with high accuracy these challenging core designs. The presented research comprises the development and implementation of the thermal–hydraulic feedback to the Monte Carlo method and of speed-up mechanisms to accelerate the Monte Carlo criticality calculation. Coupled Monte-Carlo calculations can serve as reference solutions for verifying high-fidelity coupled deterministic neutron transport methods with detailed and accurate thermal–hydraulic models. The development and verification of such reference high-fidelity coupled multi-physics scheme is performed at the Pennsylvania State University (PSU) in cooperation with AREVA, AREVA NP GmbH in Erlangen, Germany, on the basis of MCNP5, NEM, NJOY and COBRA-TF (CTF) computer codes. This paper presents the latest studies and ameliorations developed to this coupled hybrid system, which includes a new methodology for generation and interpolation of Temperature-Dependent Thermal Scattering Cross Section Libraries for MCNP5, a comparison between sub-channel approaches, and acceleration schemes.}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Espel, Federico Puente and Avramova, Maria N. and Ivanov, Kostadin N. and Misu, Stefan}, year={2013}, month={Jan}, pages={18–26} }
 @article{salko_avramova_2013, title={Uncertainty analysis of sub-channel code calculated ONB wall superheat in rod bundle experiments using the GRS methodology}, volume={65}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/J.PNUCENE.2013.02.003}, DOI={10.1016/J.PNUCENE.2013.02.003}, abstractNote={Rod bundle experiments were performed for prototypical PWR operating conditions in the project “New Experimental Studies of Thermal-Hydraulics of Rod Bundles (NESTOR)”. The intent of the project was to improve the understanding of the Axial Offset Anomaly (AOA) through improved modeling of Onset of Nucleate Boiling (ONB) (EPRI, 2008) using sub-channel codes. Skewing of the axial power profile (AOA) is most likely driven by the deposition of boron in the crud layer on nuclear fuel rods, which is caused by boiling on the fuel rod surface (EPRI, 2008). VIPRE-I (Srikantiah, 1992), a sub-channel code, was chosen for the analysis of NESTOR tests and for which uncertainty analysis was performed. NESTOR experimental results were used to optimize grid-loss coefficients, friction-loss coefficients, and a single-phase heat transfer model in the code. By modeling NESTOR ONB tests, the VIPRE-I calculated wall superheat was determined at the experimental ONB locations. This calculated ONB wall superheat could be used as a criterion in VIPRE-I for the prediction of ONB; however, it is important to quantify the uncertainty of this calculated ONB wall superheat in order to know the accuracy of such a criterion. The VIPRE-I model optimization process, however, was a complicated one and involved interaction of both experimental and code modeling uncertainties. The propagation of these uncertainties was treated using the Gesellschaft für Anlagen und Reaktorsicherheit (GRS) methodology; a process which is detailed in this paper.}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Salko, Robert K. and Avramova, Maria N.}, year={2013}, month={May}, pages={42–49} }
 @article{avramova_ivanov_2010, title={Verification, validation and uncertainty quantification in multi-physics modeling for nuclear reactor design and safety analysis}, volume={52}, ISSN={0149-1970}, url={http://dx.doi.org/10.1016/j.pnucene.2010.03.009}, DOI={10.1016/j.pnucene.2010.03.009}, abstractNote={The qualification procedure of coupled multi-physics code systems is based on the qualification framework (verification and validation) of separate physics models/codes, and includes in addition Verification and Validation (V&V) of the coupling methodologies of the different physics models. The extended verification procedure involves testing the functionality, the data exchange between different physics models, and their coupling, which is designed to model combined effects determined by the interaction of models. The extended validation procedure compares the predictions from coupled multi-physics code systems to available measured data and reference results. It is important to emphasize that such validation should be based on a multi-level approach similar to the one utilized in validating coupled neutronics/thermal–hydraulics codes in international standard problems. Appropriate benchmarks have been developed in international co-operation led by the Nuclear Energy Agency (NEA) of the Organization for Economic Cooperation and Development (OECD) that permits testing the neutronics/thermal–hydraulics coupling, and verifying the capability of the coupled codes to analyze complex transients with coupled core/plant interactions. This paper describes the above-mentioned multi-level V&V approach along with examples of the OECD benchmarks. In recent years there has been an increasing demand from nuclear research, industry, safety, and regulation for best estimate predictions to be provided with their confidence bounds. The ongoing OECD Light Water Reactor (LWR) Uncertainty Analysis in Modeling (UAM) benchmark activities contribute to establishing an unified framework to estimate safety margins supplemented by Uncertainty Quantification (UQ), which would provide more realistic, complete and logical measure of reactor safety. The paper describes the progress of the OECD LWR UAM benchmark. This activity is designed to address current regulation needs and issues related to practical implementation of risk informed regulation. Establishing such internationally accepted LWR UAM benchmark framework offers the possibility to accelerate the licensing process when using best estimate methods.}, number={7}, journal={Progress in Nuclear Energy}, publisher={Elsevier BV}, author={Avramova, Maria N. and Ivanov, Kostadin N.}, year={2010}, month={Sep}, pages={601–614} }
 @article{ivanov_avramova_2007, title={Challenges in coupled thermal–hydraulics and neutronics simulations for LWR safety analysis}, volume={34}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2007.02.016}, DOI={10.1016/j.anucene.2007.02.016}, abstractNote={The simulation of nuclear power plant accident conditions requires three-dimensional (3D) modeling of the reactor core to ensure a realistic description of physical phenomena. The operational flexibility of Light Water Reactor (LWR) plants can be improved by utilizing accurate 3D coupled neutronics/thermal–hydraulics calculations for safety margins evaluations. There are certain requirements to the coupling of thermal–hydraulic system codes and neutron-kinetics codes that ought to be considered. The objective of these requirements is to provide accurate solutions in a reasonable amount of CPU time in coupled simulations of detailed operational transient and accident scenarios. These requirements are met by the development and implementation of six basic components of the coupling methodologies: ways of coupling (internal or external coupling); coupling approach (integration algorithm or parallel processing); spatial mesh overlays; coupled time-step algorithms; coupling numerics (explicit, semi-implicit and implicit schemes); and coupled convergence schemes. These principles of the coupled simulations are discussed in details along with the scientific issues associated with the development of appropriate neutron cross-section libraries for coupled code transient modeling. The current trends in LWR nuclear power generation and regulation as well as the design of next generation LWR reactor concepts along with the continuing computer technology progress stimulate further development of these coupled code systems. These efforts have been focused towards extending the analysis capabilities as well as refining the scale and level of detail of the coupling. This article analyses the coupled phenomena and modeling challenges on both global (assembly-wise) and local (pin-wise) levels. The issues related to the consistent qualification of coupled code systems as well as their application to different types of LWR transients are presented. Finally, the advances in numerical and computation techniques for coupled code simulations are summarized with outlining remaining challenges.}, number={6}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Ivanov, Kostadin and Avramova, Maria}, year={2007}, month={Jun}, pages={501–513} }
 @article{cuervo_avramova_ivanov_miró_2006, title={Evaluation and enhancement of COBRA-TF efficiency for LWR calculations}, volume={33}, ISSN={0306-4549}, url={http://dx.doi.org/10.1016/j.anucene.2006.03.011}, DOI={10.1016/j.anucene.2006.03.011}, abstractNote={Abstract Detailed representations of the reactor core generate computational meshes with a high number of cells where the fluid dynamics equations must be solved. An exhaustive analysis of the CPU times needed by the thermal-hydraulic subchannel code COBRA-TF for different stages in the solution process has revealed that the solution of the linear system of pressure equations is the most time consuming process. To improve code efficiency two optimized matrix solvers, Super LU library and Krylov non-stationary iterative methods have been implemented in the code and their performance has been tested using a suite of five test cases. The results of performed comparative analyses have demonstrated that for large cases, the implementation of the Bi-Conjugate Gradient Stabilized (Bi-CGSTAB) Krylov method combined with the incomplete LU factorization with dual truncation strategy (ILUT) pre-conditioner reduced the time used by the code for the solution of the pressure matrix by a factor of 20. Both new solvers converge smoothly regardless of the nature of simulated cases and the mesh structures and improve the stability and accuracy of results compared to the classic Gauss–Seidel iterative method. The obtained results indicate that the direct inversion method is the best option for small cases.}, number={9}, journal={Annals of Nuclear Energy}, publisher={Elsevier BV}, author={Cuervo, Diana and Avramova, Maria and Ivanov, Kostadin and Miró, Rafael}, year={2006}, month={Jun}, pages={837–847} }