171 results on '"Mosunova, N. A."'
Search Results
2. Validation of the severe accident module of the EVKLID/V2 integral code on the base of experiments with fission products release and dissociation of nitride fuel
3. Numerical Study of Thermal Destruction of Nitride Fuel Rods Using the Severe Accident Module of the Integral Euclid/V2 Code
4. Requirements for the EUCLID-F Integral Code for the Deterministic Analysis of Accidents in Fusion Reactors
5. Application of the EUCLID Integrated Code’s HYDRA-IBRAE/LM Thermal Hydraulic Module for Analyzing the Steam Generators of Sodium Cooled Reactor Plants
6. Simulation of Melt Behavior in the Sodium-Cooled Reactor Core Catcher Using the EUCLID/V2 Integrated Computer Code HEFEST-FR Module
7. Validation of the EUCLID/V1 integrated computer code against the BOR-60 experimental data
8. Experimental Investigation of the High-Temperature Interaction of Steel with a Lead Coolant
9. Virtual-Digital Model of NFC Closure for a Fast Reactor
10. Simulating the Thermal Interaction between Fuel and Sodium Coolant Using the EUCLID/V2 Integrated Code
11. Three-Liquid, Two-Phase, Per-Channel Thermohydraulics Code Subchannel-Na/V1.0 for Single-Phase Sodium-Coolant Flows and Heat-Transfer: Key Models
12. Simulating the Behavior of Fission Product Aerosols in the Containment
13. Computational Aspects of Core Neutronics Calculation of Sodium-Cooled Fast Reactor at the Core Destruction Stage
14. Description of Models of Sodium Combustion on Premises of an NPP with a Fast Reactor Unit using the EUCLID/V2 Integrated Code and the Results of Their Validation
15. Results of Verification and Validation of the Oxid Module of the Euclid/V2 Integral Code in Part of Physicochemical Models of Processes in a Lead Coolant
16. Development of the Water Radiolysis Model in the AERMOD Module
17. Validation of the EUCLID/V1 integrated computer code against the BOR-60 experimental data
18. Verification of the EUCLID/V2 Integrated Code Thermal-Hydraulic Module Based on Experiments That Take into Account the Parameter Distribution over the Fuel Assembly’s Cross Section
19. Implementation of Parallel-Computing Algorithms for Systems with Distributed Memory in the EUKLID/V1 Integrated Computer Code
20. A Model of the Fission Products Release from the Melt Pool during Severe Accident in a Liquid Metal Cooled Fast Reactor
21. Modeling of Gas-Phase Transport in Heavy Liquid-Metal Coolant Flow in the HYDRA-IBRAE/LM Thermohydraulic Code
22. Verification of the Aerosol/Lm Module in Experiments on Sodium Burning in Moist Air
23. 3D EVKLID/V2 Code Aided Simulation of Severe Accidents
24. Verification of the EUCLID/V2 Code Based on Experiments Involving Destruction of a Liquid Metal Cooled Reactor’s Core Components
25. The EUCLID/V2 Code Physical Models for Calculating Fuel Rod and Core Failures in a Liquid Metal Cooled Reactor
26. Verification of Analytical Test Based Thermohydraulic Systems Codes for One- and Two-Phase Liquid-Metal Flows
27. The EUCLID/V1 Integrated Code for Safety Assessment of Liquid Metal Cooled Fast Reactors. Part 2: Validation and Verification
28. Experiment-Based Verification of the SAFR/V1 Module of the EVKLID/V2 Integral Code for Thermal Breakdown of Fuel Pins in a Fast Reactor
29. Experimental Simulation of Hydrodynamics and Heat Transfer in Bubble and Slug Flow Regimes in a Heavy Liquid Metal
30. AEROSOL-LM/Na Aided Simulation of Fission Product Production and Transport in the First Loop of a Fast Reactor
31. SAFR/V1 (EVKLID/V2 Integral Code Module) Aided Simulation of Melt Movement Along the Surface of a Fuel Element in a Fast Reactor During a Serious Accident
32. Fuel Pin Melting in a Fast Reactor and Melt Solidification: Simulation Using the SAFR/V1 Module of the EVKLID/V2 Integral Code
33. Experimental investigation of the impulse gas injection into liquid and the use of experimental data for verification of the HYDRA-IBRAE/LM thermohydraulic code
34. HYDRA-IBRAE/LM/V1 Thermohydraulic Code Verification Based on BN-600 Experiments
35. System of closing relations of a two-fluid model for the HYDRA-IBRAE/LM/V1 code for calculation of sodium boiling in channels of power equipment
36. Model of vapor slug growth in the channels of power engineering equipment with sodium coolant
37. Modeling of Oxide Layer Formation and Corrosion Products Coagulation and Transport in Lead Coolant Using the OXID Module of the HYDRA-IBRAE/LM Code
38. Next Generation Design Codes for a New Technological Platform for Nuclear Power
39. Fundamentals, current state of the development of, and prospects for further improvement of the new-generation thermal-hydraulic computational HYDRA-IBRAE/LM code for simulation of fast reactor systems
40. A Step in the Verification of the Hydra-Ibrae/LM/V1 Thermohydraulic Code for Calculating Sodium Coolant Flow in Fuel-Rod Assemblies
41. Recommendations on adopting the values and correlations for calculating the thermophysical and kinetic properties of liquid lead
42. Material property database
43. Implementation of Parallel-Computing Algorithms for Systems with Distributed Memory in the EUKLID/V1 Integrated Computer Code
44. Comparison of different methods used in integral codes to model coagulation of aerosols
45. Molecular Ion Mechanism of HI Formation in the First Loop
46. Recommendations on selecting the closing relations for calculating friction pressure drop in the loops of nuclear power stations equipped with VVER reactors
47. Description and simulation of physical effects in chiral guiding structures
48. The finite element method for calculating the propagation constant in a rectangular chiral waveguide
49. Application of the methodology of readiness levels for the lean development of digital twins of complex engineering systems
50. Codes of new generation - current state of development and prospects for further development
Catalog
Books, media, physical & digital resources
Discovery Service for Jio Institute Digital Library
For full access to our library's resources, please sign in.