"Thermal hydraulic behavior" trang 1 - tải miễn phí từ tailieunhanh
Sensitivity analysis of thermal-hydraulic parameters to study the corrosion intensity in nuclear power plant steam generators

Sensitivity analysis of thermal-hydraulic parameters to study the corrosion intensity in nuclear power plant steam generators
8    102    0
Multivariate analysis of critical parameters influencing the reliability of thermal-hydraulic passive safety system

Multivariate analysis of critical parameters influencing the reliability of thermal-hydraulic passive safety system
9    108    0
Temperature effect on the thermal and hydraulic conductivity of Korean bentonite buffer material

Temperature effect on the thermal and hydraulic conductivity of Korean bentonite buffer material
8    9    1
THERMAL-HYDRAULIC IN NUCLEAR REACTOR

THERMAL-HYDRAULIC IN NUCLEAR REACTOR
259    65    0
Thermal hydraulic analysis of core flow bypass in a typical research reactor

Thermal hydraulic analysis of core flow bypass in a typical research reactor
6    79    0
Thermal-hydraulic study of air-cooled passive decay heat removal system for APRþ under extended station blackout

Thermal-hydraulic study of air-cooled passive decay heat removal system for APRþ under extended station blackout
13    73    0
Thermal hydraulic simulations of the Angra 2 PWR

Thermal hydraulic simulations of the Angra 2 PWR
8    77    0
Neutronics characterization of an erbia fully poisoned PWR assembly by means of the APOLLO2 code

Neutronics characterization of an erbia fully poisoned PWR assembly by means of the APOLLO2 code
12    64    0
Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code

Thermal-hydraulics analysis for VVR-KN fuel lead test using PLTEMP code
7    77    2
Safety analysis methodology for aged candu nuclear reactors

Safety analysis methodology for aged candu nuclear reactors
8    64    0
Uncertainty quantification of RELAP5/MOD3.3 for interfacial shear stress during small break LOCA

Uncertainty quantification of RELAP5/MOD3.3 for interfacial shear stress during small break LOCA
7    80    0
A study of different cases of VVER reactor core flooding in a large break loss of coolant accident

A study of different cases of VVER reactor core flooding in a large break loss of coolant accident
11    62    0
Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/ LSTF test 7

Comparison of CATHARE results with the experimental results of cold leg intermediate break LOCA obtained during ROSA-2/ LSTF test 7
8    81    0
A sensitivity study of physical models using in RELAP5 code based on FEBA experimental data

A sensitivity study of physical models using in RELAP5 code based on FEBA experimental data
17    57    3
On nuclear-coupled thermal-hydraulic instability analysis of Super-CriticalLight-Water-cooled-reactor (SCLWR)

On nuclear-coupled thermal-hydraulic instability analysis of Super-CriticalLight-Water-cooled-reactor (SCLWR)
20    17    1
An analysis of air-water flow phenomena due to a pipe break under sub-atmospheric pressures using TRACE

An analysis of air-water flow phenomena due to a pipe break under sub-atmospheric pressures using TRACE
11    9    1
Sub-channel/system coupled code development and its application to SCWR-FQT loop

Sub-channel/system coupled code development and its application to SCWR-FQT loop
9    14    1
Verification of a real-time interactive transient simulator for Dalat nuclear research reactor

Verification of a real-time interactive transient simulator for Dalat nuclear research reactor
8    68    3