The aim of this study is to verify if the reliability of a digital four-channel RPS under the design phase satisfies the specified target and to identify the weakness of system design and potential solutions for system reliability improvement. The event-tree/fault-tree (ET/FT), which is the method used in the current probabilistic safety assessment (PSA) framework of nuclear power plants (NPPs), was adopted to developed reliability modeling for the RPS with the Top Events defined as the system failure to generate reactor trip signal and the system generating spurious trip signal. The evaluation results indicate that the probability of the system failure on demand and the frequency of spurious trip signal generation are 1.47×10⁻⁶ with a 95 upper bound of 4.63×10⁻⁶ and 7.94×10⁻⁴/year with a 95 upper bound of 2.50×10⁻³/year, respectively. The importance and sensitivity analyses were conducted and it was found that undetected unsafe common cause failures (CCFs) of signal conditioning modules (SCMs) dominate the system reliability. Two preliminary optimization schemes relative to reducing periodic test interval and adapting two kinds of diverse SCMs were proposed. Results of the quantitive evaluation of the schemes show that neither of them could determinedly improve the system reliability to the target level. In the future, more detailed optimization analysis shall be required to determine a feasible system design optimization scheme.

Square channel is widely used in the conceptual design of water-cooled blanket of fusion reactor for cooling and providing appropriate inner temperature field for tritium breeding. Blanket is one of the most important safety components, and the thermal-hydraulic characteristics of blanket directly determine the heat transfer efficiency and safe operation of fusion reactor. Under accident conditions, the natural circulation phenomenon occurs without any mechanical devices intervention when the field forces acting on the fluid produce density gradients able to induce natural convection, which are the main heat transfer mechanisms and the important measures to mitigate the consequence of the reactor accident. For square channel in blanket, the experimental study of natural circulation heat transfer has been conducted. It can be found that natural circulation mass flow is not an independent parameter, which increases with the increase in the heat flux and with the decrease in system pressure within the experiment scope. A correlation that can correlate the experimental data with the maximum deviation of 13% is established to predict natural circulation mass flow. Simultaneously, natural circulation heat transfer is strongly affected by system pressure and heat flux. And new correlations are developed on the foundation of experimental data to predict the heat transfer coefficients of natural circulation under different conditions. Comparing the experimental results with the results of forced circulation, it can be found that the heat transfer intensity of natural circulation is lower than that of forced circulation.

Sodium-cooled reactors (SFR) have always been recognized as one of the most promising candidates for the fourth-generation nuclear systems as announced by the Generation-IV International Forum. In the design of SFR, helical wire-wrapped rod is applied to stabilize the structure of the rod bundle and enhance coolant mixing. Although there has been considerable research on SFR in computational fluid dynamics (CFD), the phenomenon of heat transfer has rarely been paid attention to. This article discovered that there exists reversed heat flux from coolant to wrapped wire, which is contrary to our usual understanding. This phenomenon has not been reported in previous CFD calculations. Hence, a solid heat conduction model is proposed to prove this phenomenon and analyze the heat transfer process. The simulation results show that the wrapping wire embedding depth, the shape of the calculation domain and the physical properties of all components have great influence on the magnitude of the reversed heat flux. The present findings will have strong influence on the temperature field and maximum value of the fuel rod as well as profound reference value for future flow calculation, especially in grid generation and treatment of the junction between the winding wire and fuel rod.

Situated at the bottom of the vacuum vessel, the divertor extracts heat and ash produced by the fusion reaction, minimizes plasma contamination, and protects the surrounding walls from thermal and neutronic loads. The vertical targets of divertor are designed to be able for up to 20 MW/m2 high heat flux. It is a great ordeal for both the material performance and the cooling ability. Critical heat flux margin is very crucial during the design of divertor. The ANSYS FLUENT is used in this paper to predict the critical heat flux on a monoblock structure with a twisted tape inside the tube. Numerical results are validated with the corresponding sets of experimental results. In this paper, CFD method used to predict critical heat flux of divertor cooling channel was first introduced. On the other hand, influence of inlet subcooling on critical heat flux is studied in detail. The inlet subcooling affect the critical heat flux much complicated for the single-side heated and swirl flow channel. Whether the influencing trend or the locations of critical heat flux occurrence are different under different inlet sub-cooling. The derivations between the simulation and experimental results were no more than 32%. This study proves the CFD tools can provide efficient help on the understanding of the critical heat flux phenomenon of complex construction.