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Background: Several studies have shown the crucial role of DNASE1L3 in regulating immune function in various diseases including Systemic Lupus Erythematosus (SLE) and cancers. However, the function and expression of DNASE1L3 in Colon Adenocarcinoma (COAD) remain obscure. The aim of this study was to explore the immune function of DNASE1L3 in COAD through a comprehensive bioinformatic analysis. Objective: To study DNASE1L3 expression in colorectal cancer through multiple databases. Use bioinformatics analysis to learn about the influence of this gene on clinical prognosis and immune function in colorectal cancer, and to explore its potential biological function. Methods: We obtained transcriptome data of COAD and normal samples from The Cancer Genome Atlas (TCGA) and identified differentially expressed (DE) mRNAs. The difference mRNA between COAD and normal samples and the difference mRNA between stage I and stage IV were respectively excavated, and then the intersection was taken. DNASE1L3 was determined the significant DE mRNA. Further, we obtained the different expressions of DNASE1L3 integrates the normal tissue data in the TCGA tumor tissue data to analyze the expression differences of 20 tumors.DNASE1L3 was further subjected to an analysis of expression in a different stage of COAD. We evaluated the influence of DNASE1L3 on clinical prognosis using Gene Expression Profiling Interactive Analysis (GEPIA) in COAD patients. To demonstrate the relationship between immune function and DNASE1L3, we investigated whether DNASE1L3 expression is related to the level of immune infiltration in COAD. We separately counted the number of neoantigens in COAD and analyzed the relationship between DNASE1L3 expression and the number of antigens. we analyzed the correlation between DNASE1L3 expression and MSI (microsatellite instability), using Spearman's rank correlation coefficient. In addition, we analysis the enrichment function of DNASE1L3.Finally, we performed immunohistochemical analysis on tissue microarray of colon adenocarcinoma. Results: In this study, we observed significantly down-regulated expression of DNASE1L3 in many different cancers including COAD, which also correlated with grade. Low expression of DNASE1L3 was significantly correlated with poorer overall survival (OS) in COAD (OS HR = 0.4, P=0.032). Low expression of DNASE1L3 was found with a poor prognosis. DNASE1L3 expression was positively correlated with infiltrating levels of CD4+ T and CD8+ T cells, B cells, macrophages, neutrophils and dendritic cells (DCs) in COAD. DNASE1L3 expression showed strong correlations with diverse immune marker sets in COAD. Enrichment functional analysis revealed that DNASE1L3 was associated with immunoglobulin complexes, antibacterial humoral responses, and humoral immune responses. Conclusions: These findings suggest that DNASE1L3, which functions as a tumor suppressor gene in COAD, might be a potential therapeutic target. It is correlated with prognosis and immune infiltrating levels, including those of CD4+ T and CD8+ T cells, B cells, macrophages, neutrophils and dendritic cells (DCs) in COAD patients. These findings suggest that DNASE1L3 can be used as a prognostic biomarker for determining prognosis and immune infiltration in COAD. These findings suggest that DNASE1L3 can be used as a prognostic biomarker for determining prognosis and immune infiltration in COAD.

The Brayton cycle using helium gas as a working medium has the advantages of high conversion efficiency, simple design, compact structure and high inherent safety. It can be combined with fossil energy, nuclear energy, solar energy and other forms of heat sources, and has broad prospects for development. The Helium Brayton cycle (HBC) is a highly integrated system. The dynamic characteristics of the Brayton cycle system are complex and difficult to obtain. It is necessary to establish the mechanism model of the HBC for research and analysis, and provide the basis for subsequent control system design. Based on the equations of conservation of mass, energy, momentum and energy balance, the thermo-hydraulic models of heat accumulator, precooler, intercooler and pipeline are established in this paper. The model was established according to the characteristic curves of helium turbine, low-pressure compressor and high-pressure compressor, and the shaft speed model was established according to the energy balance equation. Coupled with the above models, a lumped parameter nonlinear model of the HBC is established. Based on the MATLAB/Simulink platform, typical transient conditions such as external heat source power disturbance and external load disturbance are calculated. The results show that with the decrease of heat source input power, the temperature of each equipment and the net efficiency of the circulation decrease. When the external load fluctuates, the imbalance of compressor and turbine torque will lead to the rapid change of shaft speed.

Steam generator (SG) is an important equipment of the nuclear power plant, and the stability of its liquid level affects the safe operation of the nuclear power plant. SG is a complex system with nonlinear, time-varying, nonminimum-phase, small stability margin and large time delay. In actual operation, it is difficult for classical PID control to ensure a satisfactory control performance. In this paper, the neural network methods are used to optimize the parameters of the PID controller, and a neural network controller is designed. The controller of the system consists of two components: a classical PID controller, which realizes control through a closed loop; a single-hidden-layer neural network based on the BP (back propagation) model. The neural network calculates the coefficients of the classic PID controller through matrix operations. Two weighting matrices are adjusted according to the gradient descent method to reduce the loss function and realize the training process. The control system is deployed to a SG simulation model through Simulink. The typical working conditions are simulated and investigated. The control performance is compared with that of the classical PID controller. Through analysis, it is confirmed that the neural network PID control system can meet the control requirements with fast response speed, short settling time, stable control effect under various working conditions, and strong anti-interference ability. The results prove that the neural network control has greater advantages and better application value than the classical PID controller.

Nonlinear relationships among different process variables are very common in nuclear reactors. When constructing a nonlinear model to for process monitoring, linear-based monitoring technique has been proved inefficient and problematic. Kernel principle component analysis (KPCA) combines the principle component analysis and kernel method to cope with fault diagnose in nonlinear system. Experiments are conducted using data samples from Generic Pressurized Water Reactor Simulator(GPWR) to simulate sensor fault and three types system faults. The nonlinear relationship among monitoring variables is analyzed. KPCA is used to capture the nonlinear relationships to distinguish the sensor fault and system fault, which shows the superiority of KPCA to detect a univariate fault. Subsequently, system faults are further analyzed as a classification problem. Samples imbalance is another common problem existing in diagnose domains. When training samples of one class outnumber than other class, traditional classification algorithm performance may deteriorate. Random under-sampling boosting (RUSboost), a hybrid sampling and boosting algorithm is used to alleviate the imbalance problem. The performances of classifiers by four metrics is evaluated. When compared to Adaboost (contains only boosting), tree variants (base model and ensemble bagged trees) and Gaussian SVM, RUS boosting shows significantly high accuracy and robust performance in dealing with different ratio of training samples, which shows that RUSboost is an excellent technique to cope with imbalanced data in fault diagnose for nuclear power plants.

Rod Ejection Accident (REA) is one of the major accidents which may result in the serious consequences during nuclear reactor operation. The study on REA is helpful to understand the physical characteristics and to verify the safety design of the core. The small pressurized water reactor (SPWR) adopts a relatively small amount of control rods. Therefore, when REA occurs, the positive reactivity introduced by the accident may be larger, resulting in more serious consequences. The integral design of SPWR results in different responses to the REA. Thus, it is necessary to investigate the REA of SPWR. In this paper, 3KeyMaster simulation platform, which is coupled with a thermo-hydraulic model and a three-dimensional space-time neutron dynamics model, is used to establish the reactor core model. Four single-beam control rods are selected to carry out REA, respectively. Simulations are performed, and results are collected. The responses of the SPWR are analyzed and accident procedures are obtained. Through analysis, it can be concluded that the serious consequences are caused when the R rod closest to the core, resulting in the maximum nuclear power peak. But the fuel center temperature and cladding temperature do not exceed the limit value, which proves the safety of SPWR and the rationality of core design.

Small pressurized water reactors (PWRs) have become a new trend in the current nuclear energy development due to their many advantages, such as compact equipment layout, high thermal efficiency, and strong cycle capability. Compared with large PWRs, small PWRs are designed to reduce the coolant inventory and increase the core power density, which is not good for nuclear safety. Severe accident studies on large PWRs cannot be directly applied to small PWRs. Loss of coolant accident is one of the main inducements of reactor core melting, which needs to be focused on prevention and treatment. It is of great significance for the safe operation of small PWRs to analyze and study severe accident induced by loss of coolant accident. In this paper, MELCOR is used to establish the severe accident analysis model of the primary loop system of a small PWR, and the loss of coolant accident is introduced to obtain the accident sequence from the shutdown of the reactor until the core degradation. At the same time, the core pressure, core liquid level and other key parameters are analyzed. The results show that in the case of a severe accident, compared with the large PWR, the small PWR takes a faster time for the pressure of the primary circuit and the containment pressure to reach equilibrium after the break accident occurs. The unbalanced radial power distribution causes the cladding of the 3rd ring to fail first. In the later stage of the severe accident, the melt plays a major role in heating the coolant. During the entire core degradation process, the upper fuel assemblies start to melt first, and the core does not completely collapse. The research results can provide reference for the formulation of severe accident management guidelines for small PWRs.

A liquid metal cooled reactor coupled by mixed gas of helium and xenon closed Brayton cycle is designed for space exploration purpose. The design and dynamics of the space reactor differ from those of traditional reactors. Therefore, it is necessary to develop a control system design and simulation platform to investigate its dynamic behavior, design and evaluate the control system. Two system codes are adopted and coupled to develop the platform. The reactor core system code is used to calculate the reactor kinetics and thermal-hydraulics of the coolant. The energy conversion system code is to compute the Brayton cycle and the heat exchanger between the primary and secondary loops. 3KeyMaster simulation platform is used to load the two system codes by an interface program through dynamic link library. The data exchange and time synchronization between 3KeyMaster simulation platform and two system code are performed to realize the coupling of two system code. By introducing typical disturbances, the calculation results are compared with those from the system codes to verify the interface program. The input and output parameters can be visualized by model components in 3KeyMaster simulation platform to analyze the dynamic behaviors, which is useful for the subsequent control system design.

Topaz-II is a space nuclear reactor developed by the Former Soviet Union. It is a thermal-neutron reactor with UO2 ceramics as fuel. Electricity is generated based on the principle of thermoelectric conversion. The reactivity is controlled by several groups control drums. Its lifetime can be three years and its reactivity feedback and further dynamic characteristics are varying with its lifetime. The positive reactivity feedback characteristics and different dynamic characteristics in different lifetime make it changeling to design a suitable control system for Topaz-II. All-coefficient adaptive control is a single-input single-output adaptive control method applied to a parameter unknown, linear time invariant or slowly varying object. In some specific conditions, it is also suitable for the non-linear systems. In this method, the model where the sum of coefficients is equal to one is reconstructed based on the original dynamic model. All coefficients in this model can be identified. Therefore, it is possible to realize the optimal control. Firstly, the model of Topaz-II nuclear reactor system is built. It includes point-reactor kinetics model with six groups delayed neutrons, a one-dimensional lumped transient thermal model of the nuclear fuel element and the moderator, a thermionic model of the converters, a control drum drive model. The reactivity feedbacks with fuel temperature and moderator are considered. Based on the model, the dynamic characteristics are analyzed. Then, with the nonlinear model of Topaz-II nuclear reactor system, the transfer functions are obtained and analyzed. Based on the transfer functions, the model where the sum of coefficients is equal to one is built by using the all-coefficient adaptive control method. After dealing with the parameter and analyzing the error, the all-coefficient adaptive control systems are designed. To evaluate the performance of the control systems, power maneuvering events including a 30%FP (Full Power) step change and a reactivity disturbance of 0.0065 are simulated. The responses with traditional and proposed control methods are evaluated and compared. The simulation results show that the all-coefficient adaptive control system has a better performance in settling time and overshoot compared with those with traditional PID controller.

Steam Generator (SG) is one of the main components of the power cycle in pressurized water reactor (PWR), and it is the hub of primary coolant circuit and secondary circuit, so the thermal hydraulic analysis of the SG is crucial in the system design and safety analysis of the PWR. The horizontal steam generator (HSG) is one of the main types SG in the PWR nuclear power plant (NPP), and its advantages are that it has more secondary side water capacity and good safety and reliability. The VVER-1000 is a PWR with a thermal power of 3000 MW, and has four HSGs for four loops. The RELAP5 has been used to model the VVER-1000’s HSG and performs the analysis described in this paper. The HSG tube bundle is modeled by three horizontal channels, and the steam control volumes above the heat transfer tube bundle are modeled with three volumes. The steam space is modeled as a steam separator and the steam reception shield is the dryer. The HSG secondary side downcomers are represented with a separate component to provide the power of the natural circulation. To verify the accuracy of the model, three different typical conditions are simulated. The simulation results show that the model built in this paper can correctly simulate the operation of the HSG in VVER-1000.

The micro-high temperature gas cooled reactor (Mi-HTR) uses Tri-structural iso-tropic (TRISO) particle as fuel, helium gas as coolant, and graphite as moderator and structural material. The reactor adopts complete passive residual heat removal technology and is inherent security. The usage of the Mi-HTR is megawatt-level micro nuclear power plant. Because of the using of hexagonal prism graphite core, the dynamic characteristics of the Mi-HTR are different from large scale high temperature gas cooled reactors and pressurized water reactors. Therefore, the accurate and reliable reactor dynamic characteristics analysis are of significance for the design of the Mi-HTR control system. The neutron kinetics model of the Mi-HTR is built based on the point reactor kinetics with the reactivity feedback mechanism. The core thermal-hydraulic model of the Mi-HTR is built based on the fundamental conservation laws. By coupling the two models, a lumped-parameter model of the Mi-HTR is established. In order to obtain an accurate transfer function model for subsequent control system design, the mechanism model is linearized based on perturbation theory, and then the transfer function model with different inputs and outputs is obtained through Laplace transform. The developed nonlinear model is implemented on MATLAB/Simulink. By introducing a variety of disturbances, the simulation results of transient responses are presented, which show that the static precision of the dynamic model is satisfactory, and the trend of the transient responses is consistent with the physical laws. The obtained transfer function model and the nonlinear model are introduced with the same disturbances, and the simulation results show that the responses of transfer functions are consistent with those of the nonlinear model. Therefore, the transfer function model can be utilized to control system design for the Mi-HTR.

China has recently promised to reach carbon neutral before 2060. To achieve this ambitious goal, it is necessary to substantially increase the electricity production from clean sources, such as renewable power and nuclear power, to replace the previous share of coal power generation. Most of the nuclear reactors currently in operation in China are pressurized water reactors (PWR). The nuclear steam supply system (NSSS) in PWR is a complex nonlinear system. The steam bypass system ensures the steam generator’s cooling capacity in accident conditions and is essential for safe and efficient plant operation. Because of the broad transport delay and non-linearity in the steam bypass control system, it is difficult for the currently widely used proportional-integral-derivative (PID) control to obtain satisfactory control effects. Therefore, it is necessary and meaningful to design a suitable steam bypass control system for NSSS. The investigation of the steam bypass system is performed by MATLAB & Simulink for a large-scale PWR. The nonlinear model is developed by applying the mass and energy conservation equations of fluids and the critical flow correction. The state-space multivariable models are derived by linearizing the nonlinear model at different operation points. Open-loop simulation experiments have been carried out to study the steam bypass system’s dynamic, and the state-space model simulation results are validated by comparing with the results from the nonlinear model. The open-loop results show that the overall system has a considerable transport delay. Thus, a linear active disturbance rejection controller (LADRC) is designed for the steam bypass system. This research assumes that the model and the amount of delay are nearly unknown; however, the proposed controller still performs better than the existing Ziegler-Nichols (Z-N) tuned PID controller. The LADRC has a shorter settling time than the PID controller; the control system can satisfy the design purposes.

A pressure wave propagation across the second loop of a sodium-cooled fast reactor may lead to severe damage to the pipes and the equipment due to a large leakage sodium-water reaction accident. Therefore, the pressure source and the pressure wave propagation calculation and analysis can be significant for a sodium-cooled fast reactor’s design and operation. A mathematical model with code was built to calculate and predict both the pressure source and the pressure wave propagation after the large leak sodium-water reaction accident occurred in an SFR steam generator. In the pressure source model, the sodium-water reaction produces hydrogen and squeezes sodium nearby. It is assumed that the sodium-water reaction carries out instantaneously, and a spherical-to-column bubble is assumed to grow up at the reaction zone. Under such assumptions, the pressure source model can obtain the hydrogen bubble’s volume and pressure in the steam generator. In the pressure wave propagation model, flow is assumed to be one-dimensional. With the basic theory of mass conservation, momentum conservation, and the Isentropic relation of the working fluid, the numerical solution is calculated by a grid divided by the method of characteristics, considering both precision and efficiency. However, plenty of parameters, both of the pressure source and the parallel channels, can affect the accident in different ways and scales. Considering that to practice a large leakage sodium-water reaction experiment can be somewhat difficult because of its danger, inconvenience, and high cost, it is quite essential to conduct a sensitivity analysis through the program for the purpose of facilitating the design of an SFR. Since most sodium-cooled fast reactor designs are multimodules, a parallel channel structure with multiple pressure boundaries is modeled for sensitivity analysis. Sensitivity analysis was performed on multiple sets of important parameters, including different hydrogen bubble properties, different pressure boundaries, and different equipment types. Following the sensitivity study, the code can be used for the accident analysis in a future commercial demonstration fast reactor by indicating various parameter guidance for the design and construction.

The steam generator of the sodium-cooled reactor is the barrier between the secondary and third loops. When the heat transfer pipe breaks, the water/steam will pour into the sodium and a sodium-water reaction will occur. The pressure in the secondary loop will increase and the rupture disks will burst to reduce the pressure. For the safety analysis, the maximum pressure is limited. For the long term, the pressure pulse weakens and the fluid flow tends to a quasi-steady state several seconds or minutes after leak initiation. Therefore, it is necessary to develop a model to investigate the dynamics of the secondary loop in the long term. The continuity equation, the momentum equation and the energy equation are used to derive the sodium flow model in the secondary loop. The sodium pressure and velocity are described by the one-dimensional differential equation. The lumped parameter method is applied and the differential equations are solved by the Euler method. FORTRAN language is compiled to develop the code. Critical equipment, including the steam generator, buffer tank, pump, rupture disks, and accident discharge tank are considered in the secondary loop model. The sodium velocity and pressure responses varying with time can be obtained. Compared with the data from the safety analysis report, the tendency of the sodium velocity and pressure is consistent. It is proved that the model is reasonable and effective to simulate and analyze the actual long-term effects of sodium dynamics.

NuScale is a small nuclear reactor that relies on natural circulation. Its modular production and inherent safety can not only be used to generate electricity in some remote areas, but also provide energy for water desalination and regional heating. However, the dynamic characteristics of the NuScale are different from those of the traditional PWRs because of its passive characteristic. Therefore, it is necessary to study and analyze the system dynamic characteristics of NuScale. The NuScale PWR model is established based on MATLAB&Simulink. It includes point-reactor kinetics model with six groups delayed neutrons, coolant system and steam generator system. The model is established based on the conservation equations of mass, energy and momentum. And the correctness of the model is verified by the comparison between the steady-state calculation results and the design parameters. Transient calculation results are verified by mechanism analysis. To evaluate the dynamic characteristics of NuScale, the sine function changes with different frequencies in reactor reactivity and feedwater mass flow rate are introduced. The amplitude and phase responses of reactor power, secondary loop steam pressure, secondary loop mass flow rate, secondary loop steam temperature, and coolant average temperature are recorded. Then the Bode plot can be drawn with amplitude and phase responses in different frequencies. To evaluate the NuScale dynamic characteristics, frequency domain analysis is performed.

The heat pipe cooled reactor adopts the solid-state reactor design concept and the heat is passively transferred out of the core through heat pipes. It is characterized by high inherent safety and simple operation and has broad application prospects in deep space exploration and propulsion, sea submarine navigation and exploration. The design of heat pipe cooled reactor is unique, and its dynamics are different from traditional water-cooled reactors. Therefore, it is necessary to develop its dynamic model and perform dynamic analysis, and in this paper, the study object of the heat pipe cooled reactor is the 100kW nuclear silent thermoelectric reactor (NUSTER-100). A nonlinear dynamic model is derived from the conservation equations of mass, energy and momentum. Point reactor kinetics equations are adopted. The linear dynamic model is constructed by linearization of the nonlinear model based on the disturbance theory and the transfer function is further derived applying Laplace transform. Both models including the nonlinear model and transfer function model are established on the MATLAB & Simulink simulation platform. Dynamic characteristic analysis contains time domain analysis and frequency domain analysis. For the time domain analysis, by introducing a variety of boundary condition disturbances, the results were compared with those from transfer function. The results are consistent and can correctly reflect the dynamic characteristics of the heat pipe cooled reactor. Therefore, the transfer function model can be applied to the subsequent design of the heat pipe cooled reactor power control system. For the dynamic analysis, it is divided into time domain and frequency domain. The time domain is to observe the change of core power and sodium temperature by introducing reactivity disturbance. For the frequency domain, after drawing the Bode plot of the transfer function, the system’s characteristics at different frequencies are analyzed. In addition, it can provide a theoretical basis for the design of the heat pipe cooled reactor power control system.

Due to the small size, the high-power density and the strong environmental adaptability, the heat pipe cooled reactor has a broad application prospect in the fields of space exploration, underwater vehicle power system and other fields[1]. It will be one of the subversive technologies which will change the future nuclear power development. The safe operation of the heat pipe cooled reactor requires an effective power control system. 3KeyMaster is a comprehensive real-time simulation analysis platform and the control system can be built by using its logic module. The interface program can be developed in its task program, and other program can be integrated in this platform[2]. The system program of the heat pipe cooled reactor has compiled as a dynamic linking library. Through the task program of the 3KeyMaster simulation platform to call the dynamic linking library by using C++ program, the system program is coupled with the 3KeyMaster simulation platform. The modules of 3KeyMaster platform exchange data with the system program by shared memory. Through the responses of the temperature due to step change in the reactor power obtained by 3KeyMaster, the transfer functions are obtained by using the system identification module in MATLAB & Simulink. The power disturbance is introduced, and simulation is carried out on the 3KeyMaster. The changes of the important parameters are analyzed to obtain its dynamic characteristics. Based on the dynamic analysis on the system responses, the control strategy for the power control system is proposed. The power control system is implemented in MATLAB & Simulink. The control system parameters are tuned to determine a group of suitable control parameters[3]. Finally, the control performance of power control system is evaluated by simulations.

The Sodium-cooled Fast Reactor (SFR) has become the research hotspot and been under development rapidly because of its small size and high power density in recent years. SFR contains a total of three circuits, primary sodium circuit, second sodium circuit, and third water circuit. The third circuit is feedwater supply system to cool the temperature of sodium in the second circuit during once-through steam generator (OTSG). Therefore, OTSG is considered as one of the most important equipment, which is essential to normal operation and safety assurance. In this article, the dynamic model of OTSG is built by the movable and fixed boundary method based on the conservation equations. Besides, the other components are also modeled to build the third water circuit model, such as the paralleling channel model, feedwater pumps model, and high pressure feedwater heaters model etc. In the MATLAB/Simulink environment, the third water circuit simulation platform is developed. The steady state condition and some typical disturbance conditions are simulated to verify the provided model. According to the dynamic characteristics and control requirements, the feedforward & feedback control system is designed to adjust the flowrate in time to insure the outlet temperature of sodium in secondary circuit within the prescribe limits. To verify the effectiveness of the provided control system, some typical conditions are simulated. The simulation results showed that the developed model can simulate the steady state and transient characteristics correctly, and the provided control strategy can achieve the control objective of a stable outlet temperature of sodium in secondary circuit. The research could provide the reference for the study of the control and operation for the OTSG and feedwater supply system in SFR.

The purpose of this paper is to demonstrate a transfer function development process applicable to pressurized water reactors, specifically for the HPR1000. Firstly, the modeling methods using conservation equations are adopted to obtain the nonlinear dynamic model of the reactor coolant system, the pressurizer, the U-tube steam generator, and the steam bypass system. Then, the linearization of each dynamic model is performed based on the small perturbation theory. Laplace transformation is carried out to the linearized dynamic model to the transfer functions. Finally, transient simulations with the nonlinear models are executed by introducing typical disturbances to verify the transfer functions. The simulation results show good agreement and prove the proposed transfer function development method. Therefore, the developed transfer function can predict the dynamic behaviors of HPR1000 and be used for control studies. The methodology is a reference to other similar reactors for transfer function development.

Nuclear power plants mostly use the multimodule once-through steam generator (OTSG) parallel system as heat exchange equipment, but the current research mostly focuses on a single OTSG. The huge system brings great challenges to modeling and control system design. In this paper, the ODE method is used to model the main equipment of a multimodule parallel system based on mechanism. The results show that the maximum error is 3.77%. Based on the characteristics of parallel system, a new control strategy is proposed and compared with other control strategies by introducing four typical disturbances. The results show that the proposed feedwater control scheme can effectively solve the coordination between loop valve and module valve of multimodule parallel OTSG system, and the coordination between main feedwater pump and loop valve. Compared with Method 1, the response time of Method 2 in sodium temperature disturbance is shortened by 319 s and the temperature fluctuation is reduced by 1.67 times, but Method 2 has potential safety hazards. Method 1 can improve the system response speed when the sodium flowrate and water flowrate change. Compared with Method 3, the outlet sodium temperature fluctuation of Method 1 is reduced by 10 times and the adjustment time of Method 1 shortens 141.9 s in the case of feedwater flow disturbance.

A sodium-cooled fast reactor (SFR) is one of Generation IV nuclear systems. The once-through steam generator is important for heat transfer and acts as a barrier to separate the sodium and water. If the heat transfer tube breaks, a sodium–water reaction (SWR) occurs, and the integrity of the secondary loop may be destroyed. Thus, a dynamic model to analyze the SWR must be developed, especially for a large-leak SWR. A large-leak SWR analysis model was developed for an SFR, consisting of four modules: water/steam leakage rate, hydrogen bubble growth, pressure wave propagation, and SWR protection system action. A large-leak SWR accident was simulated using this model, and the process was investigated. Sensitivity analysis was performed for the critical parameters, and the parameters with a key influence on the large-leak SWR were determined. The results are significant for the design of the steam generator and the SWR protection system.

The power control of nuclear reactors is the key to the safe operation of nuclear reactors. At present, PID controllers are still used in most industrial control. Although PID has the advantages of simple design and easy parameter tuning, it is still difficult to overcome its own shortcomings in the control of complex systems and large-scale parameter adjustment. Model-Free active disturbance rejection control(MF-ADRC) and PID are control method that do not require model information. This study compares the control performance of PID and MF-ADRC in nuclear power control. First, three different core models are established, and the power MF-ADRC and PID controller of LBFR is designed. The performance of MF-ADRC and PID is compared through reactive disturbance and wide-range power adjustment. Finally, the performance of MF-ADRC and PID control is compared through the control of different core power. The results show that MF-ADRC has the same parameters to be tuned as PID; PID is difficult to balance overshoot and settling time, but MF-ADRC does a good job in this aspect; PID and MF-ADRC have good robustness, In terms of results, MF-ADRC is more satisfactory.

Canadian Supercritical Water-cooled Reactor (SCWR) is a pressure-tube type SCWR under development in Canada. Once-through direct cycle is adopted in the design. The reactor and the turbine are tightly coupled. A suitable strategy is necessary to deal with the relationship between the reactor and the turbine for safe and smooth plant operation. A turbine dynamic model is derived together with the reactor model to complete the nonlinear plant model. Three basic control strategies: reactor-leading, turbine-leading and feedwater-leading are applied to the Canadian SCWR. Input-output pairings of the control systems are determined based on the different strategies and Proportional and Integral (PI) controllers are tuned to meet the control requirements. A typical load pattern is introduced to test the control performance. Through simulation, it is found that the load following capacity with the reactor-leading strategy is low while the fluctuations in the thermal parameters are large with the turbine-leading and feedwater-leading strategies. Therefore, a new control strategy is needed to deal with these drawbacks. A coordinated control strategy is proposed to achieve high load following capacity with small variation in the thermal parameters. It is established based on the turbine-leading strategy by adding a feedforwad control system to balance the reactor and the turbine. In the feedforward control system, the input signal is the difference between the steam pressure and its setpoint, and the outputs are the control rod reactivity and the feedwater flow rate. In such a way, the power-to-flow ratio can be maintained in a correct proportion to reduce the fluctuations in the thermal parameters. Simulations are performed by introducing a typical load pattern. From the responses, it is found that the turbine power can closely follow the load demand and the variation in the steam temperature is efficiently suppressed compared to that with the turbine-leading strategy. It can be concluded that the proposed control strategy is applicable for the Canadian SCWR and smooth operation can be achieved.

The Canadian SuperCritical Water-cooled Reactor (SCWR) is a once-through pressure tube–type SCWR under development in Canada. It is a multivariable system with strong cross coupling and a high degree of nonlinearity. The outputs are sensitive to disturbances, and the variations in the thermal parameters should be limited to avoid thermal stress to its components. Therefore, designing an adequate control system is challenging. In this paper, robust multivariable feedback control and feedforward control are proposed to design the control system of the Canadian SCWR. Three uncertainty sources are considered: unmodeled uncertainty, linearization uncertainty, and model reduction uncertainty. These uncertainties are evaluated taking into account all aspects affecting the linear dynamic model used in the robust controller synthesis, and the uncertainty bounds are determined to cover the uncertainties. The robust feedback controller is synthesized using the μ-synthesis approach. The feedforward control is...

Significant progress has been made in the development of the small pressurized water reactors (SPWR). Unique characteristics of the SPWR deliver challenges to its control system design. In order to facilitate the control system design process and enhance its efficiency, it is important and necessary to establish a control system design and validation platform. Using shared memory technology, an engineering simulator coupled with MATLAB/Simulink is employed to achieve this objective. Shared memory is an efficient method to exchange data within programs. Dynamic data exchange and simulation time synchronization methods are particularly treated. To verify the platform, an SPWR with its control system is modeled using the platform and the simulator. Thermal-hydraulic modeling of the SPWR is carried out using Relap5, and its nodalization is introduced. The objectives of the control strategy are to maintain the average coolant temperature linearly varying with the reactor power and steam pressure constant. A preliminary SPWR control system is designed with proportional-integral-derivative (PID) controllers, and is implemented in MATLAB/Simulink associated with the engineering simulator. Subsequently, in order to evaluate the performance of the established simulation platform, transients of abrupt load changes and wide range load changes are simulated and simulation results are verified against those obtained from the engineering simulator alone. It is demonstrated that simulation results of both platforms are consistent with each other, which proves that the coupling of engineering simulator and MATLAB/Simulink is successful. Therefore, the established simulation platform can be committed to control system design and validation. Furthermore, it is revealed that the designed control strategy is able to regulate the process parameters such as coolant average temperature and steam pressure to the desired values.

The Canadian direct-cycle Supercritical Water-cooled Reactor (SCWR) is a pressure-tube type SCWR under development in Canada. The dynamics of the steam temperature have a high degree of nonlinearity and are highly sensitive to reactor power disturbances. Traditional gain scheduling control cannot theoretically guarantee stability for all operating regions. The control performance can also be deteriorated when the controllers are switched. In this paper, a linear parameter-varying (LPV) strategy is proposed to solve such problems. Jacobian linearization and curve fitting are applied to derive the LPV model, which is verified using a nonlinear dynamic model and determined to be sufficiently accurate for control studies. An output feedback H ∞ controller is synthesized to stabilize the steam temperature system and reject reactor power disturbances. The LPV steam temperature controller is implemented using a nonlinear dynamic model, and step changes in the setpoints and typical load patterns are carried out in the testing process. It is demonstrated through numerical simulation that the LPV controller not only stabilizes the steam temperature under different disturbances but also efficiently rejects reactor power disturbances and suppresses the steam temperature variation at different power levels. The LPV approach is effective in solving control problems of the steam temperature in the Canadian SCWR.

Frequent and wide-range load changes in small pressurized water reactors applied for nuclear-powered devices are necessary under complex and harsh conditions. However, the traditional power regulation method does not consider the stroke of the control rod, which seriously affects the service life of the control rod drive mechanism. This paper proposes a data-driven reactor power regulation system (RPRS) based on the hybrid feedforward and feedback control theory. Different adjustment regions are set in the controller to adjust the proportion of the feedforward and feedback signals dynamically. The polynomial fitting method and particle swarm algorithm are applied to optimize the size of the different adjustment regions. Typical load change conditions are applied to test and verify that the data-driven RPRS can efficiently shorten the overshoot of the control rod position and improve the performance of the power regulation system. The robustness of the data-driven RPRS is also demonstrated by simulation.

The sodium-heated steam generator plays a crucial role in the operation and safety of sodium-cooled fast reactor. The configuration with multiple steam generators poses challenges to the feedwater control system design. It is necessary to adopt an appropriate control strategy. Based on the dynamic analysis, hybrid feedback and feedforward control theory is applied in the loop feedwater control. The feedback control consists of a temperature controller and a feedwater flow controller. The feedforward controller is applied to reduce the effects on the sodium temperature from the sodium flow. To eliminate the unbalance among different branches, branch temperature controller is applied. The controller parameters are tuned using trial and error method. The control performance is evaluated through simulation under symmetrical and unsymmetrical conditions. The simulation results are analyzed. It can be found the changes in the setpoint can be quickly traced and disturbances are rejected efficiently with the designed control system.

China Fast Reactor (CFR) with 600 MW electric power is under development in China. Lots of efforts are being made to its Research and Development (R&D). The CFR600 steam generator with straight tubes consists of the evaporator and the superheater. Phase change exists in the water/steam side and the thermal–hydraulic behavior is difficult to predict. It is challenging to design such a steam generator. Therefore, a suitable thermal–hydraulic code is necessary to assist its design and analysis. A steady-state thermal–hydraulic model is derived based on the mass, energy and momentum equations. The characteristic tube is divided into four regions: subcooled region, nucleate boiling region, film boiling region and superheated region. Equal node enthalpy increment in a region is applied to enhance the numerical accuracy. The model is verified with the design results of design code SG-33 and CEFR. It is proven that the code developed can be used to design and analyze the thermal–hydraulic behavior. Sensitivity analysis is performed to investigate the influence on the heat transfer from key parameters. A preliminary design of CFR600 steam generator is presented using the developed code and reasonable results are obtained.

The sodium-heated steam generator plays a crucial role in the operation and safety of sodium-cooled fast reactor. The configuration with multiple steam generators poses challenges to the feedwater control system design. Therefore, it is necessary to develop an appropriate dynamic model and investigate its dynamics. The dynamic model is derived by applying mass, energy and momentum conservation equations and lumped-parameter approach. Steady-state calculations were performed and the results agreed well with those obtained from a thermal-hydraulic design code and the design parameters. Transient simulations were carried out by introducing step disturbances to the inputs. The system responses were analyzed and the dynamics were investigated. Unsymmetrical conditions were also simulated to study the interaction responses of different channels. Through analysis, the results were reasonable in theory. Therefore, the developed model can predict the dynamics of the feedwater system with multiple steam generators and can be used for feedwater control system design.

A new multi-purpose modular small pressurized water reactor with once-through steam generators is being designed in China. Its key parameters are different from traditional large pressurized water reactor. There are sixteen once-through steam generators divided into two groups inside of the pressure vessel. The four coolant pumps are located on the periphery of the pressure vessel. The coolant is heated by the core and transported the heat to the secondary loop by once-through steam generators. The superheated steam is generated, and its dynamics are different from those of U-tube steam generators. The relationship between the reactor and turbine is also complicated and needs to investigate. The control strategies of traditional large pressurized water reactor cannot be applied directly to the small reactor with once-through steam generators. Therefore, it is necessary to investigate suitable control strategies of the multi-purpose modular small reactor with once-through steam generators. Three control strategies are proposed and investigated in this study: turbine-leading, reactor-leading and feedwater-leading. With the reactor-leading strategy, the reactor power is adjusted by moving the control rod. The coolant temperature follows the change of the reactor power. Feedwater flow is applied to regulate the steam pressure. The steam flow rate follows the change of the feedwater flow rate to satisfy the demand power. With the turbine-leading strategy, the steam valve is adjusted which will influence the steam flow to satisfy the demand power. The feedwater-leading control strategy is adjusting the feed water flow rate corresponding to the demand power which has been measured. And reactor power and turbine load vary with feedwater flow rate. Input-output pairings of the control systems are determined based on the different strategies and proportion-integral-derivative (PID) controllers are tuned to meet the control requirements. To evaluate the performance of control strategies, power maneuvering events including a 10%FP (Full Power) step change and a ramp change with a rate of 5%FP/min are simulated. The processes of important control parameters varying with time are compared and evaluated to obtain the suitable one. Conclusions can be drawn from the simulation analyses of the control performance. The reactor-leading control strategy is best for the base-load operation. The turbine-leading control strategy is more suitable for load-following operation. The feedwater leading control strategy can be applied to load-following operation with smooth load adjustment.

Small Pressurized Water Reactors (SPWR) are different from those of the commercial large Pressurized Water Reactors (PWRs). There are no hot legs and cold legs between the reactor core and the steam generators like in the PWR. The coolant inventory is in a large amount. The inertia of the coolant is large and it takes a long time for the primary system to respond to disturbances. Once-through steam generator is adopted and its water inventory is small. It is very sensitive to disturbances. These unique characteristics challenge the control system design of an SPWR. Relap5 is used to model an SPWR. In the reactor power control system, both the reactor power and the coolant average temperature are regulated by the control rod reactivity. In the feedwater flow control system, the coordination between the reactor and the turbine is considered and coolant average temperature is adopted as one measurable disturbance to balance them. The coolant pressure is adjusted based on the heaters and spray in the pressurizer. The water level in the pressurizer is controlled by the charging flow. Transient simulations are carried out to evaluate the control system performance. When the reactor is perturbed, the reactor can be stabilized under the control system.

Temperature variation in natural environment was conducted in order to predict the action spectrum of temperature under natural environment by substituting the model to the segmented fitting method. A novel method based on the measured data was proposed to solve the concrete thermal parameter. Results showed that the temperature variation of the concrete and natural environment periodically fluctuates, and the theoretical action and response spectra of the temperature agree well with the measured results. The presence or absence of sheltered condition has a significant effect on temperature response in concrete in terms of the amplitude, extreme value and arising time.

Canadian supercritical water-cooled reactor (SCWR) can be modelled as a Multiple-input Multiple-output (MIMO) system. It has a high power-to-flow ratio, strong cross-coupling and high degree of nonlinearity in its dynamic characteristics. Among the outputs, the steam temperature is strongly affected by the reactor power and the most challenging to control. It is difficult to adopt a traditional control system design methodology to obtain a control system with satisfactory performance. In this paper, feedforward control is applied to reduce the effect on steam temperature from the reactor power. Single-input Single-output (SISO) feedback controllers are synthesized in the frequency domain. Using the feedforward controller, the steam temperature variation due to disturbances at the reactor power has been significantly suppressed. The control system can effectively maintain the overall system stability and regulate the plant around a specified operating condition. To deal with the nonlinearities, gain scheduling control strategy is adopted. Different sets of controllers combined by smooth weight functions are used for the plant at different load conditions. The proposed control strategies have been evaluated under various operating scenarios. Simulation results show that satisfactory performance can successfully achieved by the designed control system.

An innovative reliability analysis approach known as “Subset Simulation based on Importance Sampling” is developed for the efficient estimation of the small functional failure probability of a passive safety system. This approach is based on the idea that a small failure probability can be expressed as a product of larger conditional failure probabilities by introducing a proper choice of intermediate failure events. Importance sampling simulation is carried out to generate conditional samples for each intermediate failure region. This application is illustrated for the functional reliability analysis of a passive residual heat removal system due to epistemic uncertainty parameters. The numerical results demonstrate the high level of computational efficiency and excellent computational accuracy by comparison with direct Monte Carlo simulation, Importance Sampling simulation and Subset Simulation based on Markov Chain Monte Carlo. The sensitivity, defined as the partial derivative of the failure probability with respect to the distribution parameter is also discussed, which can help to identify the contribution of each parameter and guide the optimization model.