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The bridge structure safety real-time monitoring system based on big data evaluates the bridge safety status and real-time warning under the comprehensive consideration of various factors that affect the service performance of the bridge. The system platform uses a distributed file system with high fault tolerance and a parallel data processing engine with high computational efficiency. It has high reliability, availability and storage efficiency, and is easy to expand. Furthermore, the multi-factor analysis method is used to fully explore the implicit correlation between the real-time data of the various sensors of the bridge, and the bridge service performance evaluation model is accurately established by analysing the data correlation, and real-time warning of the bridge safety status is given. In addition, the system platform uses a variety of model verification methods to evaluate the effectiveness of the model to ensure the reliability of the big data analysis method.

A comparative computational fluid dynamics (CFD) study was conducted on the three different types of pressurized water reactor (PWR) upper plenum, named TYPE 1 (support columns (SCs) and control rod guide tubes (CRGTs) with two large windows), TYPE 2 (SCs and CRGTs without windows), and TYPE 3 (two parallel perforated barrel shells and CRGTs). First, three types of upper plenum geometry information were collected, simplified, and adopted into the BORA facility, which is a 1/5 scale system of the four-loop PWR reactor. Then, the geometry, including the upper half core, upper plenum region, and hot legs, was built using the Salome platform. After that, an unsteady calculation to simulate the reactor balance operation at hot full power scenario was performed. Finally, the differences of flowrate distribution at the core outlet and temperature distribution and transverse velocity inside the hot legs with different upper plenum internals were compared. The results suggest that TYPE 1 upper plenum internals cause the largest flowrate difference at the core outlet while TYPE 3 leads to the most even distributed flowrate. The distribution and evolution pattern of the tangential velocity inside hot legs is highly dependent on the upper plenum internals. Two counter-rotating swirls exist inside the TYPE 1 hot leg and only one swirl revolving around the hog leg axis exist inside the TYPE 2 hot leg. For TYPE 3, two swirls like that of TYPE 1 rotating around the hot leg axis significantly increase the temperature homogenization speed. This research provides meaningful guidelines for the future optimization and design of advanced PWR upper plenum internal structures.

The asymmetric pressure wave caused by Loss of Coolant Accident (LOCA) transients in multi-loop type Pressurized Water Reactor (PWR) leads to great shock to the reactor core barrel. The instantaneous shock causes severe damage to the reactor vessel structure integrity. In this paper, a new method for reactor core barrel instantaneous force calculation was proposed based on the widely accepted reactor system analysis code RELAP5. The reliability of RELAP5 code for pipe blow down and pressure wave propagation simulation was assessed against Edward pipe problem firstly. Then the typical three-loop PWR system model was established utilizing Symbolic Nuclear Analysis Package (SNAP) software and the small break LOCA scenarios with different break diameters and break locations were analyzed. The maximum instantaneous asymmetric force applied on the reactor core barrel was achieved under various break conditions, providing the boundary conditions for the core barrel stress distribution and deformation calculation. Results show that the cold leg break LOCA transient leads to great instantaneous asymmetric force, while the hot leg break LOCA transient has no obvious asymmetric force generated. The core barrel instantaneous stress concentration and deformation were discovered. This research could contribute to the reliability assessment of PWR core barrel under LOCA scenarios.

Pressurized Thermal Shock (PTS) analysis for AP1000 under Small Break Loss of Coolant Accident (SBLOCA) was performed in this paper. The three-dimensional models of Direct Vessel Injection (DVI) line, Reactor Pressure Vessel (RPV) nozzles and downcomer were established. The mathematic models of three-dimensional thermal hydraulic and stress analysis were introduced. The numerical simulation of thermal-hydraulic mixing was carried out using Computational Fluid Dynamics (CFD) method under SBLOCA scenario. It was found that the temperature distribution in RPV downcomer depended on the injection velocity relative to cross-flow greatly. The RPV cooling center region was moving up with the increase of injection velocity. The temperature distribution was non-uniform along the circumferential orientation on RPV wall and the great temperature gradient was generated between the cooling center and other regions. The AP1000 RPV stress analysis was performed using Finite Element Analysis (FEA) method following thermal-hydraulic mixing study. Results show that the most critical zone was located in the DVI nozzle chamfering under SBLOCA transient. The stress was mainly induced by high temperature gradient and the maximum stress occured when the wall temperature has the largest reduce rate. This work is meaningful for the structure integrity study of AP1000 nuclear power plant.

The detailed thermal-hydraulic phenomena inside the lower plenum of Reactor Pressure Vessel (RPV) in Pressurized Water Reactor (PWR) affect the mass flow rate distributions at core inlet significantly, leading to an impact on the operation safety of nuclear reactors. From the engineering point of view, a homogeneous flow rate distribution at the core inlet is expected. Actually, series of design concepts of lower plenum internal structures were proposed for different PWRs in the world. In this paper, the three-dimensional CFD simulations with three types of internal structures in the lower plenum, TYPE 1 (Vortex suppression plate with the secondary supports), TYPE 2 (Flow distribution skirt board with square coolant holes), TYPE 3 (Ellipsoid perforated drums and supports), were performed. The influence of different internal structures on the flow characteristics in the lower plenum were achieved and analyzed. The three types of internal structures were adapted into the same experimental facility “BORA”, which is a 1/5th scale 4-loops mock-up model, to realize the single variable criterion and obtain a uniform standard during the comparison. Results indicate that the mass flow rates in central fuel assemblies are much higher than that in peripheral fuel assemblies at core inlets with TYPE 1 and TYPE 2 internal structures, while the core inlet mass flow rate distribution with TYPE 3 internal structure is more homogeneous. This research has high reference value for the future advanced PWR lower plenum internal structures optimization and design.

The high-fidelity simulation for nuclear reactor has attracted more attentions recent years and it could provide much deeper understanding of the thermal hydraulic features in the reactor pressure vessel, promoting the reactor design optimization or operation guidelines making. The temperature heterogeneity in hot legs induces the deviation of reactor core power level estimation. In order to investigate the lateral flow at the reactor upper plenum and temperature heterogeneity in the hot legs of AP1000, a multi-scale coupling method was proposed with the computational fluid dynamics (CFD) and sub-channel analysis in this paper. The sub-channel analysis code, COBRA-EN, was used to calculate the coolant temperature distribution at core outlet and provide the accurate boundary conditions for CFD simulation. A detailed CFD analysis model from the core outlet to hot leg outlet was established. The main internal components including control guide tubes with eight large opening windows on it and support columns were kept, while other little components including control rod assemblies were omitted to decrease the total mesh quantity. Compared with the temperature field at core outlet, the maximum temperature difference at the entrance of the hot leg drops 10.0 K after the strongly mixing in the upper plenum. The hotter coolant from central fuel assemblies remains at the upper part of the hot leg, while the cooler coolant from peripheral fuel assemblies stays in the lower part of the hot leg. The temperature heterogeneity and its evolvement along the hot leg were analyzed.

Plate-type reactor fuel is getting increasing attentions as it features excellent heat transfer ability and compact structure. Turbulence mixing accompanied with momentum and mass exchange is typical phenomena in the converging region downstream of parallel plane fuels. To deepen understanding turbulence mixing characteristics and obtain more benchmark data for CFD, an experimental study was conducted by 2D particle image velocimetry (2D-PIV). Ensemble average profile of velocity, vorticity, turbulence intensity, as well as Reynolds stress are analyzed, respectively. Results reveal that two kinds of merging points (mp) were found, i.e., flows from two side jets merge at y/d = 3 (mp1), and flows from middle three jets merge at y/d = 7.6(mp2). The decay of vertical velocity decreases fast primarily until y/d = 6, then slightly increases until y/d = 8.8, and finally decreases gradually. Decay of velocity magnitude of present study decreases sharply until y/d = 6.4, then decreases gradually. This tendency of five-parallel jets is similar to the result of two-parallel jets. From flow visualization, the vortex scale decreases once the vortex formed. The vorticity reaches its maximum at about y/d = 1.3 and then decreases gradually. Consistently with the indicated of velocity distribution, the vorticity distribution tends to mild with the flow developing downstream. However, the location of vortex center is affected by spanwise momentum exchange. Vortices started moving closer around the mp2 indicating a strong combination of activities. Spanwise turbulence intensity shows strong fluctuations existed around mp1 implying that the main momentum transfer happened in the merging region.

Bi-level optimization strategy collaborative optimization (CO) strategy is used for bridge system performance-based design. Optimization design process is subdivided into two levels: system level and discipline level. For system level, target function is formulated with the total direct construction cost. For discipline level, In terms of load types and structural seismic performance levels, bridge system design process is decomposed into three disciplines: carrying capacity discipline, E1 earthquake discipline and E2 earthquake discipline. Constraints to every discipline are formulated in terms of corresponding codes demands respectively. In order to examine the feasibility and efficiency of CO, both CO and mono-discipline optimization strategy taking Sequential Quadratic Programming as optimizer are applied to an example of a three span simple supported bridge. The results show that CO has high computing efficiency when achieving the same optimization results, especially for design process involving time-consuming disciplinary analysis process.

The intelligent technique is becoming the most important and powerful aided tool to manage the infrastructure information and reduce the urban natural hazards. This paper presents the development in building the smart digital China programs and the application of the digital technique to mitigate natural hazard in Chinese cities. The digital techniques in this paper mean the "3S" technique, which is geographical information system (GIS), globe positioning system (GPS) and remote sensing (RS). As one of the powerful and the most popular tools, 3S is applied to these systems for infrastructure information management and natural hazard reduction. So far, most of Chinese cities have built their own intelligent digital natural hazard systems, which are mainly built based on 3S. Some details techniques in the procedure of building the digital system, including the integration methods, data transferring and data communication among 3S systems, are discussed. The principle, design criteria, structure and functions of these systems are also described in this paper. As an illustration, several successful cases are illustrated.