The inherent law between fatigue behaviours of shear-type representative volume element and mode-II fatigue crack growth is found in the range of cycle plastic zone near the crack tip. Prediction models for mode-II fa...
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Stress intensity factors for penetrate radial crack at center of keyway and at keyway corner under the plane stress, the plane strain and the three dimensional, based on weight function method, is determined by finite...
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Floating nuclear power plant (FNPP) is built on the offshore platform and it is defined as a kind of Small and Modular reactor(SMR) due to its power level. The FNPP can provide power supply, fresh water and high tempe...
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With the rapid development of china's economy, the relative shortage of energy has become one of the important factors restricting economic and social development. At present, the research and development of energ...
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The elasto-plastic analysis for reactor Pressure Vessel seal is necessary because of nuclear safety is actively demanded. The existed work based on simplified way to simulate the seal ring, such as uniform stress, spr...
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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.
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