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Among the six gen-IV reactor concepts recommended by the gen-IV international forum (GIF), supercritical water-cooled reactor (SCWR), the only reactor with water as coolant, achieves a high thermal efficiency and, subsequently, has economic advantages over the existing reactors due to its high outlet temperature. A thermal-hydraulic analysis of the SCWR assembly is performed in this paper using the modified COBRA-IV code. Two approaches to reduce the hot channel factor are investigated: decreasing the moderator mass flow and increasing the thermal resistance between moderator channel and its adjacent sub-channels. It is shown that heat transfer deterioration cannot be avoided in SCWR fuel assembly. It is, therefore, highly required to calculate the cladding temperature accurately and to preserve the fuel rod cladding integrity under heat transfer deterioration conditions. __________ Translated from Nuclear Power Engineering, 2007, 28(5): 18–21, 58 [译自: 核动力工程]  相似文献
2.
The cathode flow-field design of a polymer electrolyte membrane (PEM) fuel cell is crucial to its performance, because it determines the distribution of reactants and the removal of liquid water from the fuel cell. In this study, the cathode flow-field of a parallel flow-field PEM fuel cell was optimized using a sub-channel. The main-channel was fed with moist air, whereas the sub-channel was fed with dry air. The influences of the sub-channel flow rate (SFR, the amount of air from the sub-channel inlet as a percentage of the total cathode flow rate) and the inlet positions (SIP, where the sub-channel inlets were placed along the cathode channel) on fuel cell performance were numerically evaluated using a three-dimensional, two-phase fuel cell model. The results indicated that the SFR and SIP had significant impacts on the distribution of the feed air, removal of liquid water, and fuel cell performance. It was found that when the SIP was located at about 30% along the length of the channel from main-channel inlet and the SFR was about 70%, the PEM fuel cell exhibited much better performance than seen with a conventional design.  相似文献
3.
Pb‐Bi‐cooled direct contact boiling water fast reactor (PBWFR) featured with a direct‐contact heat exchanger between lead‐bismuth eutectic coolant and water could significantly simplify the primary system and enhance the natural circulation capability, meeting the potential needs for small modular reactor design. It is of great importance to conduct thermal‐hydraulic analysis of the PBWFR core in detail. In this paper, a self‐developed SUB‐channel AnalysiS code SUBAS is adopted to study the thermal hydraulic characteristics of the PBWFR core. The fidelity and the reliability of the code have been preliminarily benchmarked. With SUBAS, the space grid is studied to figure out its impact on the temperature and flow distributions in each sub‐channel. Besides, the application of space grids would increase the pressure drops and decreases the cross flow between adjacent sub‐channels. To study the transient performance of the PBWFR core, the power transient and the inlet blockage accident are calculated by SUBAS. The results of the power transient show the cross‐flow effect would be weakened in the sub‐channel which has higher coolant temperature and larger mass flow rate. For the inlet blockage accident, the results indicate the influence of the small area blockage is relatively weak on the overall performance of the assembly but is significant on the local parameters. With consideration of time and space, the blockage influence only exists in a certain area. This research may provide contribution to the design of PBWFR.  相似文献
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