Kostadin Nikolov Ivanov

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

APPLIED SCIENCES-BASEL, 14(17).

Clustering and uncertainty analysis to improve the machine learning-based predictions of SAFARI-1 control follower assembly axial neutron flux profiles

ANNALS OF NUCLEAR ENERGY, 206.

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.

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

FRONTIERS IN ENERGY RESEARCH, 11.

Prediction and uncertainty quantification of SAFARI-1 axial neutron flux profiles with neural networks

ANNALS OF NUCLEAR ENERGY, 188.

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

EPJ NUCLEAR SCIENCES & TECHNOLOGIES, 9.

A hybrid neutronics method with novel fission diffusion synthetic acceleration for criticality calculations

NUCLEAR ENGINEERING AND TECHNOLOGY, 55(4), 1428–1438.

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

ENERGIES, 15(14).

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

ANNALS OF NUCLEAR ENERGY, 177.

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

ANNALS OF NUCLEAR ENERGY, 163.

Evidence-based background for constrained uncertainty quantification in a core transient analysis

ANNALS OF NUCLEAR ENERGY, 164.

Methodology for Discontinuity Factors Generation for Simplified P-3 Solver Based on Nodal Expansion Formulation

ENERGIES, 14(20).

Propagation of VHTRC manufacturing uncertainties with RAVEN/PHISICS

ANNALS OF NUCLEAR ENERGY, 165.

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

NUCLEAR ENGINEERING AND DESIGN, 384.

Energy group search engine based on surrogate models constructed with the RAVEN/NEWT/PHISICS sequence

NUCLEAR ENGINEERING AND DESIGN, 356.

Quantification and propagation of neutronics uncertainties of the Kozloduy-6 VVER-1000 fuel assembly using SCALE 6.2.1 within the NEA/OECD benchmark for uncertainty analysis in modelling of LWRs

ANNALS OF NUCLEAR ENERGY, 133, 732–749.

Selected papers from the 2018 Best Estimate Plus Uncertainty International Conference (BEPU 2018) Foreword

Petruzzi, A., Ivanov, K., & Ivanov, E. (2019, December 2). NUCLEAR TECHNOLOGY, Vol. 205, pp. III-IV.

Uncertainty Quantification and Propagation of Multiphysics Simulation of the Pressurized Water Reactor Core

Nuclear Technology, 205(12), 1618–1637.

Comparison of Best-Estimate Plus Uncertainty and Conservative Methodologies for a PWR MSLB Analysis Using a Coupled 3-D Neutron-Kinetics/Thermal-Hydraulic Code

Nuclear Technology, 198(2), 193–201.

OECD/NEA benchmark for time-dependent neutron transport calculations without spatial homogenization

NUCLEAR ENGINEERING AND DESIGN, 317, 177–189.

The IAEA CRP on HTGR uncertainties: Sensitivity study of PHISICS/RELAP5-3D MHTGR-350 core calculations using various SCALE/NEWT cross-section sets for Ex. II-1a

Nuclear Engineering and Design, 329, 156–166.

Advances in the Pennsylvania State University NEM code

ANNALS OF NUCLEAR ENERGY, 94, 251–262.

Benchmarking the NEM real-time core model for VVER-1000 simulator application - asymmetric core

Proceedings of the 24th International Conference on Nuclear Engineering, 2016, vol 5.

Benchmarking the Real-Time Core Model for VVER-1000 Simulator Application on Asymmetric Core Load

JOURNAL OF NUCLEAR ENGINEERING AND RADIATION SCIENCE, 3(3).

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

ANNALS OF NUCLEAR ENERGY, 96, 1–11.

New techniques for designing the initial and reload cores with constant long cycle lengths

Annals of Nuclear Energy, 99, 165–173.

Rod Internal Pressure Distribution and Uncertainty Analysis Using FRAPCON

NUCLEAR TECHNOLOGY, 197(1), 47–63.

The IAEA coordinated research programme on HTGR uncertainty analysis: Phase I status and Ex. I-1 prismatic reference results

Bostelmann, F., Strydom, G., Reitsma, F., & Ivanov, K. (2016, September). NUCLEAR ENGINEERING AND DESIGN, Vol. 306, pp. 77–88.

Code improvement and model validation for Ascó-II Nuclear Power Plant model using a coupled 3D neutron kinetics/thermal–hydraulic code

Annals of Nuclear Energy, 87, 366–374.

Examination and refinement of fine energy group structure for high temperature reactor analysis

Annals of Nuclear Energy, 80, 279–292.

Large-scale Monte Carlo neutron transport calculations with thermal hydraulic feedback

Annals of Nuclear Energy, 84, 204–219.

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

Annals of Nuclear Energy, 84, 178–196.

A cycle-specific cross-section update for real-time simulation of VVER-1000 core

Progress in Nuclear Energy, 74, 222–231.

ASSESSMENT OF MICROX-2 CODE WITH NEW ENDF/B-VII RELEASE 0 MASTER LIBRARIES

NUCLEAR TECHNOLOGY, 186(3), 305–316.

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

Annals of Nuclear Energy, 73, 122–130.

Development of an iterative diffusion-transport method based on MICROX-2 cross section libraries

ANNALS OF NUCLEAR ENERGY, 77, 335–342.

High-Fidelity Coupled Monte-Carlo/Thermal-Hydraulics Calculations

In D. Caruge, C. Calvin, C. M. Diop, F. Malvagi, & J.-C. Trama (Eds.), SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo.

Internal multi-scale multi-physics coupled system for high fidelity simulation of light water reactors

Annals of Nuclear Energy, 66, 104–112.

Iterative Transport-Diffusion Methodology For LWR Core Analysis

In D. Caruge, C. Calvin, C. M. Diop, F. Malvagi, & J.-C. Trama (Eds.), SNA + MC 2013 - Joint International Conference on Supercomputing in Nuclear Applications + Monte Carlo.

The effect of changing enrichments on core performance

Annals of Nuclear Energy, 75, 139–145.

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.

Analysis of the OECD/NRC Oskarshamn-2 BWR stability benchmark

Annals of Nuclear Energy, 67, 4–12.

High fidelity simulation of conventional and innovative LWR with the coupled Monte-Carlo thermal-hydraulic system MCNP-SUBCHANFLOW

Nuclear Engineering and Design, 262, 264–275.

Multi-group formulation of the temperature-dependent resonance scattering model and its impact on reactor core parameters

Annals of Nuclear Energy, 63, 751–762.

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.

Progress in the development of an implicit steady state solution in the coupled code TRACE/PARCS

Progress in Nuclear Energy, 66, 1–12.

Self-shielding models of MICROX-2 code: Review and updates

ANNALS OF NUCLEAR ENERGY, 64, 256–263.

DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part I: Development, implementation and verification

Annals of Nuclear Energy, 48, 108–122.

DYNSUB: A high fidelity coupled code system for the evaluation of local safety parameters – Part II: Comparison of different temporal schemes

Annals of Nuclear Energy, 48, 123–129.

Understanding using the Haling Power Depletion (HPD) as a guide for designing PWR cores

Annals of Nuclear Energy, 53, 120–128.

Deterministic modeling of higher angular moments of resonant neutron scattering

Annals of Nuclear Energy, 38(10), 2291–2297.

MCOR – Monte Carlo depletion code for reference LWR calculations

Annals of Nuclear Energy, 38(4), 731–741.

Application of Global Sensitivity Analysis to Nuclear Reactor Calculations

Procedia - Social and Behavioral Sciences, 2(6), 7726–7727.

Spectral zone selection methodology for pebble bed reactors

Annals of Nuclear Energy, 38(1), 80–87.

Transport model based on three-dimensional cross-section generation for TRIGA core analysis

Annals of Nuclear Energy, 37(9), 1254–1260.

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

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

Application of TRACE/PARCS to BWR stability analysis

Annals of Nuclear Energy, 36(3), 317–323.

Evaluation of PBMR control rod worth using full three-dimensional deterministic transport methods

Annals of Nuclear Energy, 35(6), 1050–1055.

Accuracy evaluation of pin exposure calculations in current LWR core design codes

Annals of Nuclear Energy, 35(3), 414–424.

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

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

Development of a practical Monte Carlo based fuel management system for the Penn State University Breazeale Research Reactor (PSBR)

Annals of Nuclear Energy, 35(3), 539–551.

New genetic algorithms (GA) to optimize PWR reactors

Annals of Nuclear Energy, 35(1), 113–120.

New genetic algorithms (GA) to optimize PWR reactors

Annals of Nuclear Energy, 35(1), 93–112.

New genetic algorithms (GA) to optimize PWR reactors

Annals of Nuclear Energy, 35(1), 121–131.

The three-dimensional neutron kinetics coupled with thermal-hydraulics in RBMK accident analysis

Nuclear Engineering and Design, 238(4), 1002–1025.

Utilization of two-dimensional deterministic transport methods for analysis of pebble-bed reactors

Annals of Nuclear Energy, 34(5), 396–405.

Analysis of the VVER1000 coolant trip benchmark using the coupled RELAP5/PARCS code

Progress in Nuclear Energy, 48(8), 806–819.

Application of genetic algorithms to optimize burnable poison placement in pressurized water reactors

Annals of Nuclear Energy, 33(5), 446–456.

Comparative RELAP5-3D analysis in support of the NPP DBA analysis in Ukraine

Progress in Nuclear Energy, 48(8), 891–911.

Development of enriched Gd-155 and Gd-157 burnable poison designs for a PWR core

Annals of Nuclear Energy, 33(5), 439–445.

Evaluation and enhancement of COBRA-TF efficiency for LWR calculations

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

Impact of cross-section generation procedures on the simulation of the VVER-1000 pump startup experiment in the OECD/DOE/CEA V1000CT benchmark by coupled 3D thermal-hydraulics/neutron kinetics models

Progress in Nuclear Energy, 48(8), 746–763.

Introduction to the special issue on the OECD/DOE/CEA V1000CT-1 benchmark

Progress in Nuclear Energy, 48(8), 727.

Modeling and optimization of existing beam port facility of PSBR

Annals of Nuclear Energy, 33(17-18), 1391–1395.

OECD/DOE/CEA VVER-1000 coolant transient (V1000CT) benchmark – A consistent approach for assessing coupled codes for RIA analysis

Progress in Nuclear Energy, 48(8), 728–745.

Sensitivity study on determining an efficient set of fuel assembly parameters in training data for designing of neural networks in hybrid genetic algorithms

Annals of Nuclear Energy, 33(5), 457–465.

Simulation of the VVER-1000 pump start-up experiment of the OECD V1000CT benchmark with CATHARE and TRAC-PF1

Progress in Nuclear Energy, 48(8), 922–936.

The PBMR steady-state and coupled kinetics core thermal-hydraulics benchmark test problems

Nuclear Engineering and Design, 236(5-6), 657–668.

Helios, current coupling collision probability method, applied for solving the NEA C5G7 MOX benchmark

Progress in Nuclear Energy, 45(2-4), 119–124.

Development of the VVER core loading optimization system

Annals of Nuclear Energy, 31(7), 747–772.

Minimizing the gamma radiation effects to spent fuel pool walls

Annals of Nuclear Energy, 31(5), 459–480.