"Pressurized water nuclear reactors using constant axial offset strategy" trang 1 - tải miễn phí từ tailieunhanh
Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy

Nonlinear control for core power of pressurized water nuclear reactors using constant axial offset strategy
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Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless steel under operating conditions of a pressurized water reactor

Fatigue crack growth characteristics of nitrogen-alloyed type 347 stainless steel under operating conditions of a pressurized water reactor
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Validation of computational fluid dynamics calculation using rossendorf coolant mixing model flow measurements in primary loop of coolant in a pressurized water reactor model

Validation of computational fluid dynamics calculation using rossendorf coolant mixing model flow measurements in primary loop of coolant in a pressurized water reactor model
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Numerical prediction of a flashing flow of saturated water at high pressure

Numerical prediction of a flashing flow of saturated water at high pressure
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A review of existing SuperCritical Water reactor concepts, safety analysis codes and safety characteristics

A review of existing SuperCritical Water reactor concepts, safety analysis codes and safety characteristics
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Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors

Evolution of the collective radiation dose of nuclear reactors from the 2nd through to the 3rd generation and 4th generation sodium-cooled fast reactors
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Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code

Simulation of reactivity-initiated accident transients on UO2-M5® fuel rods with ALCYONE V1.4 fuel performance code
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Comparison of oxide layers formed on the low-cycle fatigue crack surfaces of Alloy 690 and 316 SS tested in a simulated PWR environment

Comparison of oxide layers formed on the low-cycle fatigue crack surfaces of Alloy 690 and 316 SS tested in a simulated PWR environment
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Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs

Uncertainty analysis of ROSA/LSTF test by RELAP5 code and PKL counterpart test concerning PWR hot leg break LOCAs
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A study on the core loading pattern of the VVER-1200/V491

A study on the core loading pattern of the VVER-1200/V491
7    57    1
Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation

Contribution to the study of fission products release from nuclear fuels in severe accident conditions: effect of the pO2 on Cs, Mo and Ba speciation
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