Generally, one of the remarkable characteristics is: compared with the main pipe, the sizes of the auxiliary piping are smaller, which makes the auxiliary piping satisfy the criteria to be decoupled from the main pipi...
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ISBN:
(纸本)9784888983051
Generally, one of the remarkable characteristics is: compared with the main pipe, the sizes of the auxiliary piping are smaller, which makes the auxiliary piping satisfy the criteria to be decoupled from the main piping. So when the auxiliary pipes connected to the main piping are analyzed, there are two methods: one is that the auxiliary pipe is analyzed connected to the main piping, and the other is that the auxiliary pipe is analyzed not to be connected to the main pipe. Because the response spectrum at the location of the connection of the auxiliary pipes and the main piping is lacked, the previous choice is usually adopted. At present, the main software used for analyzing the stress of the pipes in nuclear power station is PEPS, but when the PEPS is adopted to analyze the small pipes, if the small pipes are analyzed connected to the main pipe, the boundaries of the thermal analysis and those of the seismic analysis are different, so the two kinds of analysis cannot be carried out at the only one model. That is, when the small pipe is analyzed connected to the main piping, two models are needed for thermal analysis and the seismic analysis, and then the results of the thermal analysis and the seismic analysis are distilled separately, and the results are combined according to determinate mode, and then the combined results are evaluated according to the criterion. Consequently, not only is the model complex, but also the combination of the seismic stress and the thermal stress is not convenient in the previous method. In the latter method, the auxiliary pipe is not connected to the main piping, accordingly, not only is the modal simple, but also the combination of the seismic stress and the thermal stress is convenient. The only problem which is needed to be solved is the seismic response spectrum and the displacements, and an approach to solve the problem is presented in the article. A response spectrum calculation method is obtained based on the single oscilla
Traveling wave reactor (TWR) is an advanced nuclear power system, which can keep the total amount of fissionable nuclides constant during its lifetime through the transformation of fissionable nuclides in the reactor,...
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The reactor coolant system (RCS) is the primary system of the nuclear power plant primary cycle. Some small module reactors (SMR) applied once-through steam generator and integrated reactor vessel. The characteristic ...
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Three primary design principles, meanwhile the design features, are carried through the R&D process of Chinese Small Module reactor (ACP100): integral layout, compact design and modularized configuration. As a sub...
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ISBN:
(纸本)9784890471676
Three primary design principles, meanwhile the design features, are carried through the R&D process of Chinese Small Module reactor (ACP100): integral layout, compact design and modularized configuration. As a substantial demonstration of the above design conceptions, an upper plenum built-in pressurizer is proposed for the pressure control systemdesign for Chinese SMR. In this design, branches with large diameters connected to the primary loop (i.e. surge line, etc.) are eliminated, and this will significantly reduce the LOCA possibility due to the pressure boundary breach. This study mainly focuses on the feature of responses of the built-in pressure control system during the system transients, such as step load increase/decrease, ramp load increase/decrease and partial loss of electrical load. Initial status of the plant, assumptions, postulated transient condition and methodology of this analysis are described in the first section. Important thermal hydraulic parameters that can picture the transient characteristics of the coolant system are given in the second section. It shows that, during those transients, no reactor safety related signals or safety guard features are triggered or actived. It also shows that Nuclear/thermal power response rapidly and stably, and average coolant temperature and system pressure are welled controlled within the range of normal deviation of the set point values of safety related system. This study results get a good agreement with the operating experiment study of the build-in pressure control system.
Nuclear reactor applied on ship or floating platform recommend miniaturization and optimization of PRHRs. Due to much larger volume and heat transfer area in PRHRs heat exchanger, miniaturization and design of PRHRs n...
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Thermal Diffusion Coefficient (TDC) is one of the most important parameters in subchannel analysis codes. Abundant research results demonstrate that TDC is mainly determined by the structures of fuel assembly, especia...
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Pressure release tank is a key component of nuclear power plant safeguard. Evaluating the fluid thrust force from high pressure and high temperature flow jet under postulated accidents is an essential task for nuclear...
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In primary circuit, pump-induced pulsation (PIP) of coolant may cause fatigue failure of reactor components. In this article, a three-dimensional numerical model of PIP was established by COMSOL, theory of acoustic im...
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