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"Engineering test reactors Design and construction."
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Numerical Simulation of Nuclear Power Plant Pile Foundation Damage Under Earthquake Action
This study investigates the pile foundation of a nuclear power plant situated on medium-soft soil. It employs an improved viscoelastic artificial boundary unit to accurately simulate the boundary conditions of the calculation area. The research utilizes a constitutive model of concrete damage plasticity for the pile foundation and an equivalent linearized model for the soil layer. Through large-scale shaking table experiments and numerical simulations, we explore the internal force distribution within the nuclear power structure’s pile foundation and assess the extent of the damage. The results indicate that damage primarily occurs in the medium-soft ground, concentrating in the upper part of the pile and affecting the entire cross-section. Subsequent numerical analyses were conducted after reinforcing the soil layer around the top of the pile. The findings demonstrate that this reinforcement leads to a more uniform and rational distribution of internal forces along the pile, significantly reducing damage. Notably, there is no severe damage extending across the entire cross-section after reinforcement. This outcome highlights the potential for improving the force distribution in the pile foundations of nuclear power structures through appropriate soil layer reinforcement. The insights gained from this study provide valuable guidance for the seismic design of nuclear power structures.
Journal Article
Shaking Table Testing of a Scaled Nuclear Power Plant Structure with Base Isolation
To investigate the seismic performance and isolation effect of a high-temperature gas-cooled reactor, a 1/20 scale model including a reactor, a spent-fuel plant, and a nuclear auxiliary plant was fabricated. In addition, 220 mm lead-rubber bearings were designed and produced for use in the shaking table test, which included both isolated and nonisolated conditions. Two historical earthquake records and three artificial earthquake motions were used to input the ground motion in the tests. The results demonstrated that the seismic performance of the plant was better and that the structure was in an elastic state, under a safe shutdown earthquake event. Isolation bearings were found to effectively reduce the dominate frequency of the structure. The acceleration amplification factor of the superstructure was found to be less than 1. The isolation test results showed that the peak of the floor response spectrum at the pressure vessel support was less than 0.1 g. In the nonisolation test, the peak of the floor response spectrum was greater than 1 g. In the isolation test, the relative displacement of the structure was less than 1.1 mm, which was relatively small. The structure maintained a good isolation performance and exhibited improved safety under extreme ground motion.
Journal Article
Effects of Accident Thermal Loading on Shear Behavior of Reinforced Concrete Members
This paper presents the findings from an experimental research project comprised of six full-scale reinforced concrete (RC) beam specimens that were subjected to combination of thermal and mechanical loads. The specimens were designed to represent typical structural members in nuclear structures. These specimens were subjected to accident thermal condition followed by mechanical loading up to failure. The parameters included in the investigation were: 1) maximum accident temperature (300 and 450[degrees]F[148.9 and 232.2[degrees]C]); 2) concrete clear cover (0.75 and 1.5 in. [19 and 38.1 mm]); and 3) one- or two-sided heating. The experimental results were used to evaluate the flexural and shear stiffness and strength of the tested specimens. The results indicate that accident thermal conditions reduce the shear strength and stiffness of RC beam specimens relative to the ambient values. The nominal shear strength calculated using ACI provisions conservatively estimated the strength of most RC beam specimens at elevated temperatures, but unconservatively estimated the strength of beams with severe heating (450[degrees]F [232.2[degrees]C]) and reduced clear cover of 0.75 in. (19 mm). Keywords: flexural stiffness; nuclear power plants; out-of-plane shear; reinforced concrete; shear stiffness; shear strength tests; thermal concrete cracking; thermal gradient.
Journal Article
Design and Experimental Analysis of Seismic Isolation Bearings for Nuclear Power Plant Containment Structures
The aim of this study was to further research the mechanical properties of epoxy thick-layer rubber isolation bearings, adopting orthogonal design and efficacy coefficient methods in order to optimize the geometric dimensions and material parameters of the bearing, summarizing the influence of various factors on the overall performance indicators of the bearing, and determining the optimal plan through parameter adjustment. Through a combination of experiments and simulations, the fundamental characteristics of epoxy thick-layer rubber isolation bearings are studied to determine the influence law of vertical pressure on their horizontal stiffness, vertical stiffness, and damping ratio. The analysis results suggest that epoxy plate thick-layer rubber isolation bearing exhibits stable deformation ability and possesses distinctive damping characteristics. Furthermore, it is observed that the horizontal stiffness of these bearings gradually diminishes as the vertical pressure increases. When the shear displacement reaches 80 mm, there is a notable strengthening effect observed in the horizontal stiffness of the bearing. This strengthening phenomenon proves advantageous in preventing damage to the bearing due to excessive displacement; furthermore, it is noteworthy that the vertical stiffness and damping ratio of the bearing increased with the rise in vertical pressure.
Journal Article
Hydride-Induced Responses in the Mechanical Behavior of Zircaloy-4 Sheets
This study aimed to investigate the impact of hydrogen content, up to 1217 ppm, on the mechanical properties of Zircaloy-4, with a particular focus on the formation and impact of hydrides. Tensile specimens were tested across a range of temperatures and hydrogen concentrations. The results revealed a pronounced ductile-to-brittle transition associated with hydride formation. When the hydrogen content in the specimens ranged between 700 and 850 ppm, a ductile-to-brittle transition was observed at temperatures of 25 °C, 50 °C, and 75 °C. At 25 °C, the ultimate tensile strength (UTS) of Zircaloy-4 linearly increased as the hydrogen concentration rose from 0 to 1217 ppm H. However, at higher temperatures, the behavior of UTS became more complex, especially in the hydrogen concentration ranges of 500–850 ppm H. Elongation (EL) in the hydrided specimens was affected by both temperature and hydrogen concentration. As hydrogen concentration increased, there was a noticeable decline in uniform EL, while non-uniform EL showed even more significant reductions. Scanning electron microscopy (SEM) analysis of the fracture surfaces revealed that quasi-cleavage features became evident when the hydrogen content reached 850 ppm H, across all tested temperatures. These findings not only provide a quantitative assessment of the safety implications of Zircaloy-4 in nuclear reactor applications but also highlight the importance of the hydrogen charging process and mechanical testing in understanding its mechanical behavior.
Journal Article
The Development and Application of One Thermal–Hydraulic Program Based on ANSYS for Design of Ceramic Breeder Blanket of CFETR
Thermal–hydraulic design and analysis is an important step for blanket design to determine the temperature of the material, hydraulics parameter and radial building, especially in scope optimization phase. Traditionally, thermal–hydraulic calculation is a tedious work for designer while combining with neutronic design iteration calculation through ANSYS graphical user interface operation. One secondary development program was developed based on the commercial software ANSYS and characteristics of ceramic breeder layer-separated type blanket characteristics. When the material type, dimension and nuclear heat deposition of each layer along the blanket radial direction act as the input parameters, this program can automatically create geometric model, generate mesh, adding material properties, set boundary condition, calculate and post-process the calculation data according to user definition by calling the ICEM and Fluent software. Finally, the visual two-dimensional temperature field and the max temperature of each layer are obtained promptly. This could help easily to determine the thermal characteristics and hydraulics parameter of blanket and reduce the workload of the operators. In this paper, the programming methodology is reported and verified the reliability and efficiency of this program by employing one blanket design scheme for thermal–hydraulic calculation for Chinese Fusion Engineering Test Reactor.
Journal Article
Design of the VISTA-ITL Test Facility for an Integral Type Reactor of SMART and a Post-Test Simulation of a SBLOCA Test
To validate the performance and safety of an integral type reactor of SMART, a thermal-hydraulic integral effect test facility, VISTA-ITL, is introduced with a discussion of its scientific design characteristics. The VISTA-ITL was used extensively to assess the safety and performance of the SMART design, especially for its passive safety system such as a passive residual heat removal system, and to validate various thermal-hydraulic analysis codes. The VISTA-ITL program includes several tests on the SBLOCA, CLOF, and PRHRS performances to support a verification of the SMART design and contribute to the SMART design licensing by providing proper test data for validating the system analysis codes. A typical scenario of SBLOCA was analyzed using the MARS-KS code to assess the thermal-hydraulic similarity between the SMART design and the VISTA-ITL facility, and a posttest simulation on a SBLOCA test for the shutdown cooling system line break has been performed with the MARS-KS code to assess its simulation capability for the SBLOCA scenario of the SMART design. The SBLOCA scenario in the SMART design was well reproduced using the VISTA-ITL facility, and the measured thermal-hydraulic data were properly simulated with the MARS-KS code.
Journal Article
Unbalance Compensation of a Full Scale Test Rig Designed for HTR-10GT: A Frequency-Domain Approach Based on Iterative Learning Control
Unbalance vibrations are crucial problems in heavy rotational machinery, especially for the systems with high operation speed, like turbine machinery. For the program of 10 MW High Temperature gas-cooled Reactor with direct Gas-Turbine cycle (HTR-10GT), the rated operation speed of the turbine system is 15000 RPM which is beyond the second bending frequency. In that case, even a small residual mass will lead to large unbalance vibrations. Thus, it is of great significance to study balancing methods for the system. As the turbine rotor is designed to be suspended by active magnetic bearings (AMBs), unbalance compensation could be achieved by adequate control strategies. In the paper, unbalance compensation for the Multi-Input and Multi-Output (MIMO) active magnetic bearing (AMB) system using frequency-domain iterative learning control (ILC) is analyzed. Based on the analysis, an ILC controller for unbalance compensation of the full scale test rig, which is designed for the rotor and AMBs in HTR-10GT, is designed. Simulation results are reported which show the efficiency of the ILC controller for attenuating the unbalance vibration of the full scale test rig. This research can offer valuable design criterion for unbalance compensation of the turbine machinery in HTR-10GT.
Journal Article
Design and Development Framework of Safety-Critical Software in HTR-PM
With the development of information technology, the instrumentation and control system of nuclear power plant nowadays rely heavily on the massive and complex software to ensure the safe and efficient operation of the power plant. The improvement of the software design and development for the safety systems has been a research focus for its decisive impact on the nuclear safety. The framework of the software design and development for reactor protection system in High Temperature Gas-Cooled Reactor-Pebble bed Module was introduced in this paper. Firstly, during the design period, in addition to multichannel redundancy, grouping of protection variables and diverse 2-out-of-4 logics were adopted by different subsystems of each channel in case of common cause failure. Then a series of development characteristics together with strict software verification and validation were performed. Thirdly, during the software test period, an improved software reliability growth model based on the Goel-Okumoto model according to the analysis of fault severity was proposed to help in estimating the reliability of the software product and identifying the software release time.
Journal Article