The neutron source system is an important system equipment for the critical assembly,playing a role in starting the reactor and eliminating monitoring blind spots.The neutron source system needs two functions:one is the storage and shielding of the neutron source;the second is to safely and smoothly shuttle the neutron source between the shielding tank and the designated location of the core.Based on the requirements of the critical assembly for the neutron source system,an electromechanical system has been designed that can safely and smoothly transport neutron sources from shielding tanks to designated locations in the core.By detecting the pressure changes of the steel wire rope on the pressure sensor,real-time judgment of whether the system has malfunctioned;the position of the neutron source can be determined in real time through the encoder and the terminal switch.Experimental verification has been carried out,meeting the design requirements and successfully applied on the critical assembly,demonstrating its simple operation,high reliability,good repeatability,and ability to quickly detect and eliminate faults.It has great practical value.
In this paper,a numerical simulation method of the Lead-bismuth Helical Coil Once-through Tube Steam Generator(LHCOTSG)based on the porous medium method is proposed.The correctness of the simulation method is validated by comparison with three-dimensional numerical simulation results.Based on this method,a LHCOTSG with a given working condition was designed and studied.Under the power target of 1.5 MW,the arrangement of four-layer helical tubes has the optimal comprehensive performance.On this basis,the performance of LHCOTSG with different geometric parameters under different working conditions was evaluated by comprehensive performance evaluation index.The multivariate linear fitting method is adopted to obtain the best geometric model within the range of working conditions and geometric parameters.The highest performance was improved by 3.37%.This study provides a new method reference for numerical simulation and design optimization of LHCOTSG.
The small natural circulation lead-cooled fast reactor operates in harsh environments and is typically left unattended.In such circumstances,sensor failures can have a significant impact on the system safety.To promptly detect and diagnose sensor faults,a method combining the Principal Component Analysis(PCA)and the Support Vector Machine(SVM)is proposed.This data-driven approach eliminates the need for detailed mathematical models of the system or prior knowledge of specific states.By integrating the PCA,it efficiently reduces data dimensions.Using the dynamic model of the small lead-cooled fast reactor built in the MATLAB/Simulink,the sensor data is generated to train and establish the fault diagnosis model,and the test is carried out to verify the accuracy and effectiveness of the proposed fault diagnosis method.
Nuclear power fire protection is an important part of nuclear safety, and strengthening fire protection technology research is essential for improving nuclear safety and fire protection. The operating platform inside a containment structure is one important element in fire risk evaluation in nuclear power plants. In this paper, a combustible nuclear-grade cable in a fire scenario was firstly selected as the research object, and then the nuclear-grade cable was separately subjected to a combustion test as well as a thermogravimetric test in order to obtain the relevant thermal parameters, which provide more accurate data support for the establishment of a cable fire development and spread model. The nuclear-grade cable material data obtained from the test were compared with a civil PVC cable in order to conduct a specific analysis of the fire risk of nuclear-grade cables. This study shows that the effects of different heating rates and heating atmospheres on the thermal decomposition behavior of cable materials are meaningful and helpful in understanding whether the materials will undergo combustion processes at high temperatures and whether fire spread will occur.
A three-dimensional numerical simulation model for salt water cooling tower was established to study the characteristics of drifts and salt deposition in nuclear power project.In the model,site meteorological data,design parameters of cooling tower,general layout and environmental characteristics are considered.The results show that when the natural wind direction is parallel to the tower row,two hot air streams are superimposed on each other as one,which is beneficial to reducing the influence of drifts and salt deposition in the environment.The drifts with particle size greater than 550 μm can not escape from the outlet of cooling towers.For normal operation of 2 nuclear power units at the same site,the salt deposition will not cause damage to plants.
