"nuclear fuel for considerations in pyroprocessing safeguards" trang 3 - tải miễn phí từ tailieuXANH
On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding

On the effect of temperature on the threshold stress intensity factor of delayed hydride cracking in light water reactor fuel cladding
5    92    0
Three dimensional flow phenomena in a wire wrapped 37-pin fuel bundle for sfr

Three dimensional flow phenomena in a wire wrapped 37-pin fuel bundle for sfr
11    61    0
Monte Carlo analysis of LWR spent fuel transmutation in a fusionfission hybrid reactor system

Monte Carlo analysis of LWR spent fuel transmutation in a fusionfission hybrid reactor system
10    49    0
Investigation of molten fuel coolant interaction phenomena using real time X-ray imaging of simulated woods metal-water system

Investigation of molten fuel coolant interaction phenomena using real time X-ray imaging of simulated woods metal-water system
9    80    0
Post irradiation analyses of U-MO dispersion fuel rods of komo tests at HANARO

Post irradiation analyses of U-MO dispersion fuel rods of komo tests at HANARO
12    53    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    51    4
Partitioning and transmutation strategy R&D for nuclear spent fuel: the SACSESS and GENIORS projects

Partitioning and transmutation strategy R&D for nuclear spent fuel: the SACSESS and GENIORS projects
8    81    0
Reassessment of gadolinium odd isotopes neutron cross sections: scientific motivations and sensitivity-uncertainty analysis on LWR fuel assembly criticality calculations

Reassessment of gadolinium odd isotopes neutron cross sections: scientific motivations and sensitivity-uncertainty analysis on LWR fuel assembly criticality calculations
11    49    0
Structural integrity assessment and stress measurement of chasnupp-1 fuel assembly skeleton: under tensile loading condition

Structural integrity assessment and stress measurement of chasnupp-1 fuel assembly skeleton: under tensile loading condition
6    69    0
Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly Part A: under tensile loading condition

Structural integrity assessment and stress measurement of CHASNUPP-1 fuel assembly Part A: under tensile loading condition
6    51    0
Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors

Three-component U-Pu-Th fuel for plutonium irradiation in heavy water reactors
12    43    0
Computational burnup analysis of the TRIGA Mark II research reactor fuel

Computational burnup analysis of the TRIGA Mark II research reactor fuel
21    4    1
Modelling FeCrAl cladding thermo-mechanical performance. Part I: Steady-state conditions

Modelling FeCrAl cladding thermo-mechanical performance. Part I: Steady-state conditions
13    7    1
Feasibility of minor actinide transmutation in the VVER-1000 fuel assembly

Feasibility of minor actinide transmutation in the VVER-1000 fuel assembly
11    10    2
Natural convection heat transfer characteristics in a canister with horizontal installation of dual purpose cask for spent nuclear fuel

Natural convection heat transfer characteristics in a canister with horizontal installation of dual purpose cask for spent nuclear fuel
10    54    0
Predicting behavior of AP-1000 nuclear reactor fuel rod under steady state operating condition by using FRAPCON-4.0 software

Predicting behavior of AP-1000 nuclear reactor fuel rod under steady state operating condition by using FRAPCON-4.0 software
8    50    0
Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors

Physical and economical aspects of Pu multiple recycling on the basis of REMIX reprocessing technology in thermal reactors
7    44    0
Storage of thermal reactor fuels – Implications for the back end of the fuel cycle in the UK

Storage of thermal reactor fuels – Implications for the back end of the fuel cycle in the UK
9    49    0
Sensitivity and uncertainty analysis on reactivity for HEU and LEU fuel assemblies of dalat nuclear research reactor using monte carlo code and ENDF/B-VII.0 and ENDF/B-VII.1 nuclear libraries

Sensitivity and uncertainty analysis on reactivity for HEU and LEU fuel assemblies of dalat nuclear research reactor using monte carlo code and ENDF/B-VII.0 and ENDF/B-VII.1 nuclear libraries
14    21    1
Study of experimental core configuration of the modified STACY for measurement of criticality characteristics of fuel debris

Study of experimental core configuration of the modified STACY for measurement of criticality characteristics of fuel debris
8    7    1
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