Maria Nikolova Avramova

High-fidelity multi-physics guidelines for model validation and uncertainty quantification

Delipei, G. K., Faure, Q., Avramova, M., & Ivanov, K. (2026, February 4). Frontiers in Nuclear Engineering, Vol. 2.

BWR core thermal–hydraulic uncertainty and sensitivity analysis with improved bypass modeling features

Ziyad, D., Abarca, A., & Avramova, M. (2025, February 26). Nuclear Engineering and Design, Vol. 435.

CTF development, verification, and validation for VVER core thermal-hydraulics and multi-physics modeling and simulation

Kutlu, Y., Spasov, I., Mitkov, S., Rouxelin, P., Abarca, A., Kolev, N., … Ivanov, K. (2025, August 1). Frontiers in Nuclear Engineering, Vol. 8.

Development of Serpent/CTF External Coupling for the OECD/NEA TVA-WB1 Benchmark Activities

Albagami, T., Rouxelin, P., Abarca, A., Palmtag, S., Avramova, M., & Ivanov, K. (2025, April 25). Nuclear Science and Engineering, Vol. 4.

High-resolution predictions of the coolant properties for the 3D PWR core with artificial neural networks based on CTF

Papadionysiou, M., Delipei, G., Avramova, M., Ferroukhi, H., & Ivanov, K. (2025, July 1). Nuclear Engineering and Design, Vol. 442.

Sensitivity and uncertainty analysis in pebble-bed reactors: A study using the High-Temperature Code Package (HCP)

Yaseen, M., Sadek, A., Osman, W., Altahhan, M., Wu, X., Avramova, M., & Ivanov, K. (2025, April 10). Annals of Nuclear Energy, Vol. 219.

Uncertainty Quantification Framework for High-Temperature Gas-Cooled Reactors Using VSOP and DAKOTA

Delipei, G. K., Altahhan, M., Rouxelin, P., Sen, S., George, N., Avramova, M., & Ivanov, K. (2025, May 13). Nuclear Science and Engineering, Vol. 5.

Verification of the PWR core solver coupling the neutron code nTRACER with artificial neural networks for thermal–hydraulic feedback

Papadionysiou, M., Delipei, G., Avramova, M., Ferroukhi, H., & Ivanov, K. (2025, December 18). Frontiers in Energy Research, Vol. 12.

A Framework for Multi-Physics Modeling, Design Optimization and Uncertainty Quantification of Fast-Spectrum Liquid-Fueled Molten-Salt Reactors

Holler, D., Bhaskar, S., Delipei, G., Avramova, M., & Ivanov, K. (2024, August 28). Applied Sciences, Vol. 14.

Fuel performance code to code comparative analysis for the OECD/NEA MPCMIV benchmark

Faure, Q., Delipei, G., Scolaro, A., Avramova, M., & Ivanov, K. (2024, November 13). Nuclear Engineering and Design, Vol. 430.

Investigation of Transients with Serpent using the OECD/NEA C5G7-TD Benchmark

Fernandes, A. C. A. A., Rouxelin, P., Ivanov, K., & Avramova, M. (2024, January 1). EPJ Web of Conferences, Vol. 302.

Challenge for the validation of high-fidelity multi-physics LWR modeling and simulation: Development of new experiments in research reactors

Frontiers in Energy Research, 11.

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

Nuclear Engineering and Design, 412, 112474.

Fuel performance uncertainty quantification and sensitivity analysis in the presence of epistemic and aleatoric sources of uncertainties

Frontiers in Energy Research, 11.

Preservation of kinetics parameters generated by Monte Carlo calculations in two-step deterministic calculations

EPJ Nuclear Sciences & Technologies, 9, 15.

A Three-Dimensional Cartesian Thermal-Hydraulic Solver in the Residual Formulation of CTF

Proceedings: 2022 ANS Winter Meeting. Presented at the 2022 ANS Winter Meeting, Phoenix, AZ, USA.

Ab initio molecular dynamics investigation of γ-(U,Zr) structural and thermal properties as a function of temperature and composition

Journal of Nuclear Materials, 561, 153523.

CTF-PARCS Core Multi-Physics Computational Framework for Efficient LWR Steady-State, Depletion and Transient Uncertainty Quantification

Energies, 15(14), 5226.

CTF: A modernized, production-level, thermal hydraulic solver for the solution of industry-relevant challenge problems in pressurized water reactors

Nuclear Engineering and Design, 397, 111927.

Development of a High-Fidelity Multi-Physics Coupling Between MCNP6.2 and CTF4.0 for VVER Applications

Development of a Steady-State Solver and Non-Linear Iteration Structure in the Residual Formulation of CTF

Proceedings: The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19). Presented at the 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Brussels, Belgium.

Extending a low-order inhomogeneous adjoint equations model to a higher-order model with verification on integral applications

Annals of Nuclear Energy, 177, 109277.

Fuel Performance Uncertainty Quantification and Sensitivity Analysis Methodology for Aleatoric and Epistemic Sources of Uncertainties

Proceedings: 2022 ANS Winter Meeting. Presented at the 2022 ANS Winter Meeting, Phoenix, AZ, USA.

High-to-Low Multi-Physics Modeling and Uncertainty Propagation Using NEAMS Tools”

High-to-Low Multi-Physics Modeling and Uncertainty Propagation Using NEAMS Tools.” Proceedings: 2022 ANS Winter Meeting. Presented at the 2022 ANS Winter Meeting, Phoenix, AZ, USA.

Integral Benchmark Evaluation Of Zero-Power Tests And Multi-Cycle Depletion Experimental Data Of TVA Wb1 Cycles 1-3

Proceedings: 2022 ANS Winter Meeting. Presented at the 2022 ANS Winter Meeting, Phoenix, AZ, USA.

Introducing the OECD/NEA WPRS Task Force on Artificial Intelligence and Machine Learning

Proceedings: 2022 ANS Winter Meeting. Presented at the 2022 ANS Winter Meeting, Phoenix, AZ, USA.

Investigation of γ-(U,Zr) structural properties and its interfacial properties with liquid sodium using ab initio molecular dynamics

Journal of Nuclear Materials, 567, 153835.

Investigation of Different Planar Averaging Techniques in NEK5000 for High-to-Low Information of CTF

Proceedings: The 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19). Presented at the 19th International Topical Meeting on Nuclear Reactor Thermal Hydraulics (NURETH-19), Brussels, Belgium.

Machine Learning-Based Prediction of Departure from Nucleate Boiling Power for the PSBT Benchmark

OECD/NEA MPCMIV Benchmark -- Preliminary Fuel Performance Results

Off-line vs. semi-implicit TH-TH coupling schemes: A BEPU comparison

Annals of Nuclear Energy, 178, 109344.

Phase I Specifications and Preliminary Sensitivity Analyses of the OECD-NRC Liquid Metal Fast Reactor Core Thermal-Hydraulics Benchmark

Quantification of the Importance of the Modeling Fidelity in the Best Estimate Plus Uncertainty Analysis

Status of the OECD/NEA Watts Bar Unit 1 Benchmark and Calculations of Local Reactor Core Power Data of the Zero Power Physics Tests

Why is Uncertainty Quantification Important for Machine Learning Models?

Proceedings: 2022 ANS Winter Meeting. Presented at the 2022 ANS Winter Meeting, Phoenix, AZ, USA.

julia for Enhancing Nuclear Engineering Simulations (JENES): Introduction to the JENES Project and Platform

Calibration of Pellet Cladding Gap Heat Transfer Model for PWR, BWR and VVER Fuel Rod Transient Uncertainty Analysis

Development and verification of a higher-order mathematical adjoint nodal diffusion solver

Annals of Nuclear Energy, 163, 108548.

Development of a General MPI Coupling Interface for Multi-Physics Analysis

Development of a Thermal Hydraulic Model for the Versatile Test Reactor Using the CTF Subchannel Code

Digital-twin-based improvements to diagnosis, prognosis, strategy assessment, and discrepancy checking in a nearly autonomous management and control system

Annals of Nuclear Energy, 166, 108715.

High to Low Methodology for Gap Heat Transfer in an Uncertainty Quantification Framework for Depletion

Innovations in Multi-Physics Methods Development, Validation, and Uncertainty Quantification

Journal of Nuclear Engineering, 2(1), 44–56.

Modeling of Koeberg Power Plant: Generation of Nominal and Perturbed Nuclear Data With Polaris for PARCS Simulations

Propagation of VHTRC manufacturing uncertainties with RAVEN/PHISICS

Annals of Nuclear Energy, 165, 108667.

Status of the OECD/NEA TVA Watts Bar 1 Benchmark and Monte Carlo Modeling of the "Zero Power Physics Test" Exercise

Summary of comparative analysis and conclusions from OECD/NEA LWR-UAM benchmark Phase I

Nuclear Engineering and Design, 384, 111474.

Uncertainty Quantification Study of Fuel Rod Performance With FRAPCON/FRAPTRAN for the OECD/NEA LWR Uncertainty Analysis in Modelling Benchmark

Best-Estimate Plus Uncertainty Framework for Multiscale, Multiphysics Light Water Reactor Core Analysis

Science and Technology of Nuclear Installations, 2020, 1–18.

Development and assessment of a nearly autonomous management and control system for advanced reactors

Annals of Nuclear Energy, 150, 107861.

Modular high temperature gas reactor core modeling with RELAP5-3D/PHISICS – Optimization schemes for load following

Nuclear Engineering and Design, 362, 110526.

Preliminary design and analysis of Liquid Fuel Molten Salt Reactor using multi-physics code GeN-Foam

Nuclear Engineering and Design, 369, 110826.

A surrogate model based on sparse grid interpolation for boiling water reactor subchannel void distribution

Annals of Nuclear Energy, 131, 51–62.

CTF-R: A novel residual-based thermal hydraulic solver

Nuclear Engineering and Design, 348, 37–45.

Developments in thermal-hydraulic sub-channel modeling for whole core multi-physics simulations

Nuclear Engineering and Design, 358, 110387.

Gap conductance modeling I: Theoretical considerations for single- and multi-component gases in curvilinear coordinates

Nuclear Engineering and Design, 353, 110283.

Gap conductance modeling II: Optimized model for UO2-Zircaloy interfaces

Nuclear Engineering and Design, 355, 110289.

LWR - Reactivity Transients and Accidents

In R. P. Martin & C. Frepoli (Eds.), Design-Basis Accident Analysis Methods for Light-Water Nuclear Power Plants (pp. 2591–7862). London: World Scientific Publishing Co. Ptc. Ltd.

On the validity of the dilute gas assumption for gap conductance calculations in nuclear fuel performance codes

Nuclear Engineering and Design, 350, 1–8.

A new fuel modeling capability, CTFFuel, with a case study on the fuel thermal conductivity degradation

Nuclear Engineering and Design, 341, 248–258.

Application of sub-channel modeling to BWR core analysis

Gosdin, C., & Avramova, M. (2018, February 5). Annals of Nuclear Energy, Vol. 115, pp. 294–302.

CFD prediction of critical heat flux in vertical heated tubes with uniform and non-uniform heat flux

Nuclear Engineering and Design, 326, 403–412.

Implementation and assessment of wall friction models for LWR core analysis

Toptan, A., Porter, N. W., Salko, R. K., & Avramova, M. N. (2018, February 22). Annals of Nuclear Energy, Vol. 115, pp. 565–572.

Multiphysics Analysis of CMC Silicon Carbide and Zircaloy Cladding

Presented at the 2018 26th International Conference on Nuclear Engineering.

Quantified Validation with Uncertainty Analysis for Turbulent Single-Phase Friction Models

Nuclear Technology, 205(12), 1607–1617.

Three-dimensional multi-physics modeling of hydrogen and hydride distribution in zirconium alloy cladding

Mankosa, M., & Avramova, M. (2018, February 22). Progress in Nuclear Energy, Vol. 105, pp. 294–300.

Uncertainty Quantification Study of CTF for the OECD/NEA LWR Uncertainty Analysis in Modeling Benchmark

Nuclear Science and Engineering, 190(3), 271–286.

Validation of CTF Void Predictions Using the BFBT Database

Presented at the 2018 26th International Conference on Nuclear Engineering.

Validation of CTF pressure drop and void predictions for the NUPEC BWR database

Nuclear Engineering and Design, 337, 291–299.

Comparative analysis of different fidelity fuel performance models for fuel temperature predictions

Raja, F., Mankosa, M., & Avramova, M. (2017, July 26). Nuclear Engineering and Design, Vol. 322, pp. 464–473.

Development, verification and application of a new model for active nucleation site density in boiling systems

Nuclear Engineering and Design, 328, 1–9.

Evaluation of modeling options for in-pellet power distribution and gap gas conductance for accurate fuel temperature predictions

Raja, F., & Avramova, M. (2017, June 16). Progress in Nuclear Energy, Vol. 100, pp. 135–145.

High-fidelity multi-physics coupling for determination of hydride distribution in Zr-4 cladding

Davis, I., Courty, O., Avramova, M., & Motta, A. (2017, July 15). Annals of Nuclear Energy, Vol. 110, pp. 475–485.

Multi-physics code system with improved feedback modeling

Yilmaz, M. O., Avramova, M. N., & Andersen, J. G. M. (2017, March 8). Progress in Nuclear Energy, Vol. 98, pp. 94–108.

Multiphysics modeling of nuclear reactors

AccessScience, 10.

Coupled MCNP6/CTF code: Development, testing, and application

Bennett, A., Avramova, M., & Ivanov, K. (2016, May 28). Annals of Nuclear Energy, Vol. 96, pp. 1–11.

Development, verification, and validation of a fuel thermal conductivity degradation model in CTF

Yilmaz, M. O., Avramova, M. N., & Andersen, J. G. M. (2016, July 22). Annals of Nuclear Energy, Vol. 97, pp. 246–261.

Impact of fuel thermal conductivity degradation on Doppler feedback during rod ejection accident

Yilmaz, M. O., Avramova, M. N., & Andersen, J. G. M. (2016, August 5). Nuclear Engineering and Design, Vol. 307, pp. 339–353.

High-fidelity multi-physics system TORT-TD/CTF/FRAPTRAN for light water reactor analysis

Annals of Nuclear Energy, 84, 234–243.

Multi-physics and multi-scale benchmarking and uncertainty quantification within OECD/NEA framework

Annals of Nuclear Energy, 84, 178–196.

Coupled 3D neutronics/thermal hydraulics modeling of the SAFARI-1 MTR

Annals of Nuclear Energy, 73, 122–130.

Investigations on coupled code PWR simulations using COBRA-TF with soluble boron tracking model

Progress in Nuclear Energy, 77, 72–83.

Multi-dimensional boron transport modeling in subchannel approach: Part II. Validation of CTF boron tracking model and adding boron precipitation model

Nuclear Engineering and Design, 278, 713–722.

OECD/NEA BENCHMARK FOR UNCERTAINTY ANALYSIS IN MODELING (UAM) FOR LWRS – SUMMARY AND DISCUSSION OF NEUTRONICS CASES (PHASE I)

Nuclear Engineering and Technology, 46(3), 313–342.

Optimization and parallelization of the thermal–hydraulic subchannel code CTF for high-fidelity multi-physics applications

Annals of Nuclear Energy, 84, 122–130.

Selected Papers from OECD-NEA PSBT Benchmark

Science and Technology of Nuclear Installations, 2014, 1–3.

A multi-scale homogenization and reconstruction approach for solid material temperature calculations in prismatic high temperature reactor cores

Nuclear Engineering and Design, 256, 1–13.

Assessment of CTF Boiling Transition and Critical Heat Flux Modeling Capabilities Using the OECD/NRC BFBT and PSBT Benchmark Databases

Science and Technology of Nuclear Installations, 2013, 1–12.

Comparative Analysis of CTF and Trace Thermal-Hydraulic Codes Using OECD/NRC PSBT Benchmark Void Distribution Database

Science and Technology of Nuclear Installations, 2013, 1–12.

Multi-dimensional boron transport modeling in subchannel approach: Part I. Model selection, implementation and verification of COBRA-TF boron tracking model

Nuclear Engineering and Design, 278, 701–712.

Overview and Discussion of the OECD/NRC Benchmark Based on NUPEC PWR Subchannel and Bundle Tests

Science and Technology of Nuclear Installations, 2013, 1–20.

Uncertainty analysis of sub-channel code calculated ONB wall superheat in rod bundle experiments using the GRS methodology

Progress in Nuclear Energy, 65, 42–49.

New developments of the MCNP/CTF/NEM/NJOY code system – Monte Carlo based coupled code for high accuracy modeling

Annals of Nuclear Energy, 51, 18–26.

Analyses of Single-Phase Heat Transfer and Onset of Nucleate Boiling in Rod Bundles

Verification, validation and uncertainty quantification in multi-physics modeling for nuclear reactor design and safety analysis

Progress in Nuclear Energy, 52(7), 601–614.

Challenges in coupled thermal–hydraulics and neutronics simulations for LWR safety analysis

Annals of Nuclear Energy, 34(6), 501–513.

Evaluation and enhancement of COBRA-TF efficiency for LWR calculations

Annals of Nuclear Energy, 33(9), 837–847.

Development of Pressurized Water Reactor Integrated Safety Analysis Methodology Using Multilevel Coupling Algorithm

Nuclear Science and Engineering, 148(3), 414–425.

Multilevel Methodology in Parallel Computing Environment for Evaluating BWR Safety Parameters

Nuclear Technology, 146(3), 267–278.

Temporal adaptive algorithm for TRAC-BF1/NEM/COBRA-TF coupled calculations in BWR safety analysis

Annals of Nuclear Energy, 29(18), 2127–2141.