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Meteodiffusive Characterization of Algiers' Nuclear Research Reactor
Meteodiffusive Characterization Algiers Nuclear Research Reactor
2009/9/3
In the framework of the environmental impact studies of the nuclear research reactor of Algiers, we will present the work related to the atmospheric dispersion of releases due to the installation in n...
Thermal-Hydraulic Analysis of Coolant Flow Decrease in Fuel Channels of Smolensk-3 RBMK during GDH Blockage Event
Thermal-Hydraulic Coolant Flow
2009/9/3
One of the transients that have received considerable attention in the safety evaluation of RBMK reactors is the partial break of a group distribution header (GDH). The coolant flow rate blockage in o...
Characterization and Application of the Thermal Neutron Radiography Beam in the Egyptian Second Experimental and Training Research Reactor (ETRR-2)
Egyptian Second Experimental Training Research Reactor
2009/9/3
The Experimental, Training, Research Reactor (ETRR-2) is an open-pool multipurpose reactor (MPR) with a core power of 22 MWth cooled and moderated by light water and reflected with beryllium. It...
Approaches, Relevant Topics, and Internal Method for Uncertainty Evaluation in Predictions of Thermal-Hydraulic System Codes
Uncertainty Evaluation Thermal-Hydraulic System Codes
2009/9/3
The evaluation of uncertainty constitutes the necessary supplement of best-estimate calculations performed to understand accident scenarios in water-cooled nuclear reactors. The needs come from the im...
The present paper reviews activates relevant to the boiling water reactor (BWR) stability phenomenon, which has a coupled neutronic and thermal-hydraulic nature, from the viewpoint of model and code d...
Thermal-Hydraulic Analysis Tasks for ANAV NPPs in Support of Plant Operation and Control
Thermal-Hydraulic ANAV NPPs
2009/9/3
Thermal-hydraulic analysis tasks aimed at supporting plant operation and control of nuclear power plants are an important issue for the Asociación Nuclear Ascó-Vandellòs (ANAV). ANAV is the consortium...
Thermal-Hydraulic System Codes in Nulcear Reactor Safety and Qualification Procedures
Thermal-Hydraulic System Codes Nulcear Reactor
2009/9/3
In the last four decades, large efforts have been undertaken to provide reliable thermal-hydraulic system codes for the analyses of transients and accidents in nuclear power plants. Whereas the first ...
Reliability Assessment of 2400 MWth Gas-Cooled Fast Reactor Natural Circulation Decay Heat Removal in Pressurized Situations
2400 MWth Gas-Cooled Fast Reactor Natural Circulation
2009/9/3
As the 2400 MWth gas-cooled fast reactor concept makes use of passive safety features in combination with active safety systems, the question of natural circulation decay heat removal (NCDHR) re...
Dynamic Analysis of Coolant Channel and Its Internals of Indian 540 MWe PHWR Reactor
Coolant Channel Indian 540 MWe PHWR Reactor
2009/9/3
The horizontal coolant channel is one of the important parts of primary heat transport system in PHWR type of reactors. There are in all 392 channels in the core of Indian 540 MWe reactor. Each ...
Buckling of Imperfect Thin Cylindrical Shell under Lateral Pressure
Lateral Pressure Cylindrical Shell
2009/9/3
The strength of thin shells, under external pressure, is highly dependent by the nature of imperfection. This paper investigates buckling behaviour of imperfect thin cylindrical shells with analytical...
Natural Circulation Characteristics at Low-Pressure Conditions through PANDA Experiments and ATHLET Simulations
Natural Circulation Low-Pressure PANDA Experiments
2009/9/3
Natural circulation characteristics at low pressure/low power have been studied by performing experimental investigations and numerical simulations. The PANDA large-scale facility was used to provide ...
Flow Stagnation under Single and Two-Phase Natural Circulation Conditions in the APEX-CE Test Facility
Flow Stagnation Natural Circulation Conditions the APEX-CE Test Facility
2009/9/3
Natural circulation experiments were conducted at Oregon State University using the advanced plant experiment (APEX-CE) integral system test facility as configured to simulate a typical 2×4 Combustion...
Numerical Simulations and Design Optimization of the PHT Loop of Natural Circulation BWR
Numerical Simulations Natural Circulation BWR
2009/9/3
Mathematical modeling and numerical simulation of natural circulation boiling water reactor (NCBWR) are very important in order to study its performance for different designs and various off-design co...
Evaluation of Heat Removal from RBMK-1500 Core Using Control Rods Cooling Circuit
Heat Removal RBMK-1500 Core
2009/9/3
The Ignalina nuclear power plant is a twin unit with two RBMK-1500, graphite moderated, boiling water, multichannel reactors. After the decision was made to decommission the Ignalina NPP, Unit 1 was s...
Wastes Management Through Transmutation in an ADS Reactor
Wastes Management Transmutation ADS Reactor
2009/9/3
The main challenge in nuclear fuel cycle closure is the reduction of the potential radiotoxicity, or of the time in which that possible hazard really exists. Probably, the transmutation of minor actin...