"nuclear power engineering" trang 3 - tải miễn phí từ tailieuXANH
On - power detection of pipe wall thinned defects using ir thermography in NPPs

On - power detection of pipe wall thinned defects using ir thermography in NPPs
10    64    0
Calibration of digital wide-range neutron power measurement channel for open-pool type research reactor

Calibration of digital wide-range neutron power measurement channel for open-pool type research reactor
8    67    0
Evaluation of dynamic behavior of coagulation-flocculation using hydrous ferric oxide for removal of radioactive nuclides in wastewater

Evaluation of dynamic behavior of coagulation-flocculation using hydrous ferric oxide for removal of radioactive nuclides in wastewater
8    58    0
Control of accidental discharge of radioactive materials by filtered containment venting system: A review

Control of accidental discharge of radioactive materials by filtered containment venting system: A review
12    47    0
Prognostics for integrity of steam generator tubes using the general path model

Prognostics for integrity of steam generator tubes using the general path model
9    75    0
Development of logical structure for multi-unit probabilistic safety assessment

Development of logical structure for multi-unit probabilistic safety assessment
7    185    0
OECD/NEA study on the economics and market of small reactors

OECD/NEA study on the economics and market of small reactors
6    58    0
Dynamic assessment of the seismic isolation influence for various aircraft impact loads on the CPR1000 containment

Dynamic assessment of the seismic isolation influence for various aircraft impact loads on the CPR1000 containment
15    54    0
Concept and methodology for evaluating core damage frequency considering failure correlation at multi units and sites and its application

Concept and methodology for evaluating core damage frequency considering failure correlation at multi units and sites and its application
16    7    1
Sensitivity analysis of an Advanced Gas-cooled Reactor control rod model

Sensitivity analysis of an Advanced Gas-cooled Reactor control rod model
10    5    1
Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2 – Results of severe accident analyses for unit 3

Overview and outcomes of the OECD/NEA benchmark study of the accident at the Fukushima Daiichi NPS (BSAF), Phase 2 – Results of severe accident analyses for unit 3
12    6    2
A TakagieSugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation

A TakagieSugeno fuzzy power-distribution method for a prototypical advanced reactor considering pump degradation
9    44    0
The impact of power coefficient of reactivity on candu 6 reactors

The impact of power coefficient of reactivity on candu 6 reactors
8    61    0
Research on flow characteristics in supercritical water natural circulation: Influence of heating power distribution

Research on flow characteristics in supercritical water natural circulation: Influence of heating power distribution
9    72    0
Study of a betavoltaic battery using electroplated nickel 63 on nickel foil as a power source

Study of a betavoltaic battery using electroplated nickel 63 on nickel foil as a power source
5    41    0
Rapid and massive throughput analysis of a constant volume high-pressure gas injection system

Rapid and massive throughput analysis of a constant volume high-pressure gas injection system
7    57    0
Measurement of safety rods reactivity worth by advanced source jerk method in HWZPR

Measurement of safety rods reactivity worth by advanced source jerk method in HWZPR
5    50    0
Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor

Neutronic analysis of control rod effect on safety parameters in Tehran Research Reactor
7    49    0
Study on the structure optimization and the operation scheme design of a double tube once through steam generator

Study on the structure optimization and the operation scheme design of a double tube once through steam generator
14    44    0
A fission matrix based validation protocol for computed power distributions in the advanced test reactor

A fission matrix based validation protocol for computed power distributions in the advanced test reactor
10    12    1
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