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Based on the finite difference method, the fixed boundary method, and a staggered grid method, this paper develops a one-dimensional thermal–hydraulic analysis code to simulate the effects of heave motion and more complex oceanic motions, namely heaving coupled with inclination and heaving coupled with swaying, on the thermal–hydraulic parameters of the International Reactor Innovative and Secure (IRIS) reactor's helical coil once-through steam generator (HCOTSG). The oceanic conditions were achieved by modifying the momentum equation. The validity and accuracy of the code were obtained by comparing the results with the design parameters of HCOTSG for two types of marine nuclear reactors, the SMART reactor and the IRIS reactor, and with the calculations of some similar international prediction programs. The simulation results indicate that (1) when only heave motion is considered, increasing the amplitude and reducing the period both result in a greater magnitude and a faster rate of change in the HCOTSG parameters. (2) When the combined motion of inclining and heaving is present, the dynamic performance of the composite motion inclining towards the positive x-axis deteriorates the most severely, while its thermodynamic conditions improve slightly. In this direction, the inclining angle has little influence on the thermal–hydraulic characteristics of HCOTSG. (3) When the combined motion of heaving and swaying is present, the dynamic performance is more severely affected under swaying motion towards the negative y-axis, despite a slight improvement in thermodynamic conditions. In this direction, as the swaying angle increases, the trend of parameter variation under swaying is similar but with a greater magnitude of change. The deterioration in dynamic performance becomes more severe, although there are small improvements in thermodynamic conditions. With the swaying period increasing, the maximum and minimum values of the secondary side pressure drop show little chan, QC 20240408

This paper develops a one-dimensional thermal–hydraulic analysis program to simulate the main thermal–hydraulic parameter changes of helical coil once-through steam generator (HCOTSG) under inlet thermal–hydraulic parameters perturbation in inclining and vertical conditions. The inclining process is divided into the process of rotating from the vertical position to the inclining position and the stable inclining process. The rotating process is regarded as swaying motion. The swaying and inclining motions are achieved by modifying the momentum equation. What's more, the code was verified through experiments on two-phase heat transfer and friction pressure drop, as well as comparisons with the design parameters of the HCOTSG of the International Reactor Innovative and Secure (IRIS) reactor. The results of the simulation indicate that the direction of incline has an obvious impact on the safety of the HCOTSG, with inclination towards the y-axis having the greatest impact. In the stable inclining process, an increase in the angle of inclining results in a rise in primary-side outlet temperature and a reduction in heat transfer, while the secondary-side pressure drop increases. Furthermore, the HCOTSG underwent testing with eight types of transient perturbations, including four sudden perturbations and four linear perturbations. Results indicated that during sudden perturbations, apart from the secondary-side pressure drop, there were no substantial differences in simulated results between the inclined and vertical states at the same time during the transition process. However, during linear perturbations, due to the slow changes in parameters, the differences between the inclined and vertical states at the same time were more distinct. Regardless of the perturbation type, the inclined state led to a deterioration in the dynamic condition of the system compared to the vertical state., QC 20240215

The occurrence of porous deposits on the core surface during the operation of a pressurized water reactor can seriously affect its heat transfer characteristics and increase the surface temperature of the cladding. Continued heat transfer deterioration causes subcooled boiling and boron enrichment within porous deposits, leading to axial power anomalies in the core, and the study of porous deposits is beneficial to improving the economics and safety of reactor operation. In this paper, a two-dimensional boiling unit model of a porous deposits was established using COMSOL software to couple heat conduction, darcy flow, solute transport and chemical reactions in a multiphysics field to reasonably predict the temperature field, velocity field, boric acid concentration distribution, and PH distribution within the porous deposits. On this basis, the influence of different parameters (porosity, boiling unit radius of porous deposits, heat flow density) on the maximum effective thermal conductivity, the pattern of thermal parameters (temperature, superheat, percentage of subcooled boiling, and saturation temperature) on the chimney boundary are analyzed. The results show that localized superheating, maximum boric acid concentration and subcooled boiling occur at the bottom of the porous deposits. There is a negative correlation between the porosity and the percentage of subcooled boiling area, and a positive correlation between the radius of the porous deposits and the heat flow density and the percentage of subcooled boiling area., QC 20240912

Steam generators are essential for the safe, economical, and reliable operation of nuclear reactors. Helical Coil Once-through Steam Generators (HCOTSG) offer unique advantages over other types of steam generators for ship reactors or offshore platforms. To investigate the effect of rolling motion due to ocean conditions on the safe and economic operation of HCOTSG, a new model and a new code are developed in this paper. The program is developed and allows numerical simulation of HCOTSG under ocean conditions. The validity of the code was verified by comparing the simulation results with the design parameters of the Marine Reactor X (MRX) steam generator and the simulation results of other dedicated programs. The code was further used to perform the International Reactor Innovative and Secure (IRIS) transient operating conditions under typical rolling motions. Then, the influences of different swing directions, angles, periods, and positions of swing axes from the origin of the system operating parameters are analyzed. In the cases discussed in this paper, the following conclusions are obtained: (a) The direction of the swing significantly affects the system. The most dangerous situation is around the x-axis, and the case around the z-axis is the safest, in which the rolling situation has little effect on the system because the centripetal force is perpendicular to the tube wall, and the gravitational pressure drop is constant. (b) when the swing angle increases, except the fluctuation speed is faster, the fluctuation range of the parameters also increases. (c) when the swing period changes, the parameters’ fluctuation range also change (d) the orientation of the swing axis from the origin affects the magnitudes of the parameters’ changes (e) the distance of the swing axis from the origin affects the magnitude of the parameter variation and the farther the swing axis is from the origin, the greater the parameter fluctuation., QC 20230904

At the beginning of the new century, since nuclear energy received widespread attention due to its economy and environmental protection, the nuclear industry has developed rapidly. However, there are still some issues challenging the safety of nuclear power plants (NPPs) especially at extreme conditions. During a hypothetical severe accident in a light water reactor, the core fuel will be disrupted and melt-down resulting from the absence of coolant, and finally forms a high-temperature molten pool at the lower head of reactor pressure vessel (RPV), which threaten the integrity of RPV, one of the most important barriers to prevent radioactive fission product releasing to the environment. In this regard, the decay heat removal of the molten pool plays an essential role in the strategy of in-vessel retention (IVR)-external reactor vessel cooling (ERVC) to arrest the molten corium inside the RPV, as shown in Fig. 1.This paper examines the research work about experimental investigation, theoretical analysis, and numerical simulation conducted for ERVC and summarizes the different approaches proposed by scholars to strengthen ERVC capabilities. Those approaches are the replacement of the cooling work materials, the application of nanofluids, surface modification, and other methods. Finally, it concludes the deficiencies of the current study and the direction of future efforts., QC 20230706