INNOVATIVE DESIGNS AND TECHNOLOGIES OF NUCLEAR POWER ISTC NIKIET – 2026

Conceptual design solution for the power section of the NPP with the BREST-OD-300 reactor unit with a gas topping system

Rationale for the use of heavy-duty concrete for the BREST-OD-300 RU

Technical and economic optimization of main technical solutions for thermal-mechanical equipment of the BR-1200 turbine island

Thermohydraulic stability study of BR-1200 RP’s SG

Steam generator for large sodium-cooled fast neutron reactor plants

Modular nuclear power plant with lead-bismuth coolant and a gas turbine power converter for Arctic conditions (SVGT-1)

Experimental study of transient processes in a small-scale multi-loop circuit with heavy liquid metal coolant

On technical aspects of ensuring radiation safety of Generation IV nuclear energy systems

Development of next-generation digital twins creation technology using the example of virtual test bench complex implementation

Assessment of MBIR reactor potential to conduct advanced experimental research

Engineering conceptualization of a molten salt curium burner reactor

Physical characteristics of dual-purpose pressure-tube reactor

Production of medical isotopes at operating RBMK-1000 power units: technology, safety, and operational experience

Concept of a channel-type graphite reactor for isotope production and an evaluation of its parameters

Analysis of nuclear reactor - powered spacecraft projects to fly beyond LEO

Issues of regulation of the spacecraft safety with nuclear reactors

Criticality experiments for small-sized fast multiplying configurations with high enriched uranium nuclear fuel

Thermal energy storage based on molten salt for the BREST-OD-300 reactor

Implementation of ejector for justification of residual heat removal system’s heat exchanger operability of floating power unit reactor plant in natural circulation mode

CAD SIMINTECH experience for implementation of computational task algorithms in APCS

Justifying sites for installing primary transducers of the means for monitoring temperature and thermodynamic activity of oxygen in lead coolant of BR-1200 RP

Dependence of VVER-440 fuel load operating time on control element assemblies position and thermal power

Influence of viscosity and thermal conductivity parameters of the oxide phase of corium melt on heat transfer conditions at its boundaries under conditions of stratification of molten materials during a severe accident of a nuclear power plant

Hydrogen fuel cells for new generation PAR

Determination of the relative fission gas removal rate for the evaluation of the BREST-OD-300 fuel rod leakage

Dependence of the thermal state of a metallic melt layer under asymmetric side cooling conditions during a severe accident

Study of local flow characteristics in a heated seven-rod FA mockup with self-spacing fuel elements based on modified models

Rational arrangement of spacer grids of nuclear reactor fuel assemblies

Detection of Ag-110m in KAB and FAK systems of LNPP-2: the results of routine radiation monitoring

Features of nitrogen-15 concentration and synthesis of mixed uranium-plutonium nitride based on it

Numerical modeling of the compaction process for structural material segments from irradiated fuel assemblies

Thermodynamic analysis of a recompression S-CO₂ cycle for the SHELF-M SMR

Assessment of the influence of different control and safety system placement in the IBR-4 active zone on its characteristics

Reasonability of taking magnetic neutron scattering by electrons into account in numerical analyses of thermal nuclear reactors

Neutronic features of filling the core of a molten salt research reactor during the physical start-up stage

Nomenclature of in-reactor measurement tools and their application in nuclear research reactors of RIAR JSC

Development of the choked flow model for the next-generation computational codes

The results of pre-test calculations of experiments for loss of cooling accidents of VVER-1200 SFP modeling using KMS large scale facility

Theoretical and experimental study of post-crisis heat transfer during cooling of hot bodies in liquids

CFD modeling of the influence of biofouling on plate heat exchangers of intermediate circuits systems consumers of the reactor compartment

Thermodynamic analysis of options for applying external steam superheating at nuclear power plants

Application of neural networks for modeling interphase mass transfer of radiolytic hydrogen in the primary system equipment of a VVER-type reactor

Application of the GA-DPSA approach and machine learning methods for accident analysis of VVER nuclear power plants

Cross-platform verification and validation of thermal-hydraulic models of nuclear installations

Verification of the MCU-NR software system for validation of the SHELF-M-based SNPP design

Using machine learning with reinforcement to create reactor plant control algorithms

Development of the FRIDAY automated preparation system for multi-group neutron cross-section libraries

Simulation of the IBR-2M nuclear power plant neutronics when using a new FA set with a modified isotopic composition of plutonium

Calculation of the passive feedback device drop dynamics in a liquid lead flow using CFD simulation

SADCO SS and ACCORD-med for calculating the accumulation of medical isotope activity

CROSS-CUDA: high-speed group nuclear data constant preparation system for enhancing the accuracy of engineering neutronic calculations

Simulation of processes in an experimental mixing chamber using multidimensional CFD model integrated into the HYDRA-IBRAE/LM code

Capabilities of express non-destructive testing for monitoring susceptibility to intergranular corrosion and intergranular stress corrosion cracking using new portable measuring system

Structural behavior of steel-concrete composite structures with external steel-plate reinforcement under accidental heating

On the results of applying a simplified pump unit model for seismic fragility assessment

Oxidation-free heat treatment and leak testing of welds in absorber rods for the BREST reactor core

Stress categories in the assessment of static strength of corner welds

Justification of changes in the scope and frequency of operational non-destructive testing of the reactor vessel metal of unit 4 of the Kalinin NPP using PSA methods