Fuyu Zhao

Application of an Improved Mechanical Shim Control Strategy for AP1000 Reactor

Abstract This paper presents the performance evaluation of an improved mechanical shim (MSHIM) control strategy that is implemented in the AP1000 reactor by a digital rod control system. The MSHIM control system automatically controls the core reactivity and axial power distribution using gray and black M control banks (M-banks) and an axial offset (AO) control bank (AO-bank). The M-banks and AO-bank are independently controlled by the power control subsystem and the AO control subsystem. In the original MSHIM strategy, the power control subsystem takes precedence, and the AO-bank is blocked from moving when a demand signal exists for the movement of the M-banks. This rod control logic can minimize the potential for interactions between the two rod control subsystems and guarantee the safety and stability of the MSHIM control system. However, the AO control capability is weakened at the same time. Thus, Westinghouse has improved this core control strategy, which gives preference to the AO-bank when both the AO-bank and the M-banks have a demand to move in the same direction. In this paper, first, the coupling characteristic of the MSHIM control strategy is analyzed to illustrate the coupling effect between the two rod control subsystems. Then, both the original and the improved MSHIM control strategies are applied to AP1000. It has been demonstrated by the MSHIM load-follow and load regulation simulation results that the improved strategy not only can provide much tighter AO control but also can reduce the total control rod movement without compromising the coolant average temperature control. Therefore, the improved MSHIM strategy can provide much better reactor control capabilities than the original strategy.

Multi-Objective Optimization of Double-Tube Once-Through Steam Generator

This paper presents a double-tube once-through steam generator (DOTSG) consisting of the outer straight tube and the inner helical tube. The tube length and pressure drop of are important parameters in optimal design of DOTSG. For optimal design of such a system, it was modeled to estimate its tube length and pressure drop. Pitch of inner helical tube, flow distribution ratio of the primary fluid, and tube assemblage are considered as design parameters. Then fast and elitist nondominated sorting genetic algorithm-II (NSGA-II) method was applied to find the optimum values of design parameters. In the presented optimal design approach, the tube length and the total pressure drop are two objective functions. The results of optimal designs were a set of multiple optimum solutions, called “Pareto optimal solutions.” The sensitivity analysis of change in optimum tube length and pressure drop with change in design parameters of the DOTSG is also performed and the results are reported.

Prediction Study on PCI Failure of Reactor Fuel Based on a Radial Basis Function Neural Network

Pellet-clad interaction (PCI) is one of the major issues in fuel rod design and reactor core operation in water cooled reactors. The prediction of fuel rod failure by PCI is studied in this paper by the method of radial basis function neural network (RBFNN). The neural network is built through the analysis of the existing experimental data. It is concluded that it is a suitable way to reduce the calculation complexity. A self-organized RBFNN is used in our study, which can vary its structure dynamically in order to maintain the prediction accuracy. For the purpose of the appropriate network complexity and overall computational efficiency, the hidden neurons in the RBFNN can be changed online based on the neuron activity and mutual information. The presented method is tested by the experimental data from the reference, and the results demonstrate its effectiveness.

Modeling and Simulation of Nuclear Power Units for Primary Frequency Regulation Research

The power system frequency stability problem has been the focus of attention since 1920s. Primary frequency regulation (PFR) is one of vital methods to maintain frequency stability, especially when the power system suffers a sudden load disturbance, such as the substation tripping or generator outage accident. To a power system, the more power units participating in PFR, the frequency stability will be controlled better. With the rapid development of nuclear power units in China, the study of the nuclear power units participating in PFR of power system has becoming a pressing problem. In the present study, a detailed nonlinear dynamic mathematical model of the whole pressurized water reactor (PWR) nuclear power plant is built. The dynamic calculation codes are compiled to dynamic link library (DLL) files, which are developed by using FORTRAN language, and the compiled DLL files are embedded into MATLAB/SIMULINK simulation platform by using S-function, the control systems are designed and the whole system of PWR nuclear power plant is simulated in MATLAB/SIMULINK finally. In this method, the operation and control mode of PWR participating in PFR of power system is analyzed and simulated. The simulations results show that the PWR nuclear units are feasible in participating in PFR from safety and economy.Copyright

Analysis of Flow Instability for OTSG Using Frequency Domain Control Theory

Abstract Two-phase flow instability in parallel channels of a once-through steam generator (OTSG) is systematically analyzed by the modern frequency domain method. The mathematical expressions of heat transfer and flow for an OTSG are proposed, and the transfer function of the closed-loop system is deduced by the use of linearization and Laplace transfer. The OTSG’s stability is judged according to the Nyquist Stability Criterion. The instability of OTSG in two cases (single-channel model and multichannel model) is researched, respectively, for a numerical example. The result shows the classical frequency domain method can be commonly used when the coupling effects in the system are negligible, and it is only a special case of the multivariable method. Furthermore, the stability sensitivity to the operating parameters is analyzed for the OTSG in this paper. The predicted results are in agreement with the experimental results given in the literature.

Characteristic Optimization of the Double-Tube OTSG

The OTSG (Once-Through Steam Generator) is usually used in the integral nuclear power equipment which requires smaller size and better effect of heat transfer. The OTSG with double-side heat transfer component is presented in this paper. The heat transfer component is composed of straight tube outside and helix tube inside. In the both sides of the helix tube, the flow is spirally, therefore, the heat transfer is enhanced. The smaller the pitch, the stronger the spirally flow, the effect of heat transfer is better, but the flow resistance is raised. Especially the increased flow resistance in the secondary side brings a great influence to the pump. The heat transfer region of the secondary fluid are divided into three regions: sub-cooled region, boiling region, and superheated region, the effects of heat transfer induced by the spirally flow vary in different regions. Thus, there is an optimization problem which is to find an optimization pitch of the inner helix tube with the best effect of heat transfer and the minimum flow resistance. Based on analyzing the effects of the pitch on heat transfer enhancement and flow resistance, the pitch is optimized by the constrained nonlinear optimization method.Copyright

Transient Simulations of CPR1000 Nuclear Power Plant Implementing Advanced Mechanical Shim Control System

Abstract A novel concept of implementing the advanced mechanical shim (MSHIM) control system on the improved Chinese Pressurized Water Reactor (CPR1000) is proposed. The reactor power control system of CPR1000 is redesigned to adopt the MSHIM control system while the other parameters and control systems remain unchanged. To investigate the control performance and safety margins of this reconfiguration, the CPR1000 Full-Scope Simulation Platform (CFSSP) is first developed in MATLAB/Simulink with relevant control systems and protection system considered. The CFSSP consists of the one-dimensional nodal core model, the nonequilibrium three-region pressurizer model, the lumped-parameters dynamic model of U-tube steam generator with movable boiling boundary, and the balance of plant model. Based on the CFSSP, operational transients of step and linear turbine load changes were simulated and analyzed. The simulation results agree well with physical laws and the control performance is satisfactory. All key parameters are kept within acceptable ranges with enough safety margins and thus the protection system is not triggered. Therefore, the CPR1000 nuclear power plant implementing the MSHIM control system can safely sustain the ±10% full-power (FP) step changes and ±5% FP/min linear changes of load transients. This study can serve as a reference for the MSHIM control system application to pressurized water reactors.

Conventional Controller Design for the Reactor Power Control System of the Advanced Small Pressurized Water Reactor

Abstract The objective of this paper is to design a reactor power control system for the advanced small pressurized water reactor that adopts a constant average coolant temperature and a secondary-side steam pressure program. Based on the nonlinear core model with the one-group delayed neutron and simplified nonlinear once-through steam generator model, a two-input and two-output linear nuclear steam supply system (NSSS) model is obtained. Three types of control systems are then proposed and designed on a Bode diagram using analytical methods and second-order approximation. The comparison of the control performance and robustness of the three control systems shows that the double feedback control (DFC) with both power feedback and temperature feedback provides the best performance for reactor power and average coolant temperature with parameter uncertainty due to control rod differential worth variation. The simulations based on the high-order nonlinear NSSS model also show good performance of the DFC system.

Development of a Fast Simulation Program for AP1000 Reactor With Application of Mechanical Shim (MSHIM) Control Strategy and Nodal Method

The Mechanical Shim (MSHIM) core control strategy makes use of two independently controlled rod groups to provide fine control of both core reactivity and axial power distribution. This paper presents a reactor core fast simulation program (RCFSP) for AP1000 using MATLAB/SIMULINK. A nodal core model including xenon iodine dynamics is used to describe the core thermal power transient with the two group neutron diffusion equation for neutron kinetics modeling and an integral method for thermal-hydraulic calculation. Two closed loop rod controllers with implementation of the MSHIM core control strategy are developed to modulate the insertion of control rods. Based on the developed RCFSP, the MSHIM load follow operations with the original and revised MSHIM control strategies and two typical MSHIM load regulation operations with ten-percent step load change and five-percent per minute ramp load change are simulated. Results of these MSHIM operations demonstrate that the core reactivity and axial power distribution can be well-controlled via automatic rod control only. It has also been demonstrated that the MSHIM capabilities provided by the original MSHIM strategy are not diminished by the revised one. Moreover, the M-bank insertion for the original strategy is much deeper than that for the revised one. Thus, the power distribution perturbations associate with the M-bank movement for the revised strategy are not as pronounced as those for the original one during load change transients, which helps to alleviated peaking factor concerns associated with the control rod insertion. In view of its accuracy, simplicity and fast computation speed, the developed RCFSP can be used for dynamic simulations and control studies of the AP1000 reactor with application of MSHIM control strategy. With the adoption of modular programming techniques, the RCFSP code can be easily modified and applied to other pressurized water nuclear reactors that employs separate, independent control rod banks for respectively controlling core reactivity and axial offset within corresponding deadbands.© 2014 ASME

Modeling and Control Strategy of the China LBE Cooled Fast Reactor

China lead bismuth eutectic (LBE) cooled fast reactor comprises of the primary system with lead bismuth eutectic (LBE) as the coolant, the secondary circuit with sub-cooled water as the coolant and the associated air cooling system for an effective rejection of thermal power to the environment as a final heat sink. The dynamic characteristics of LBE cooled fast reactor are different from the traditional Pressurized Water Reactors (PWRs) because of the variances in coolant properties as well as major differences due to the operation in the fast versus the thermal neutron spectrum. To investigate the dynamic characteristics of the CLEAR-IA reactor for control system design and simulation, a model for the main components of the reactor and the most relevant interactions among them is developed. Since all the coefficients in the models are functions of temperature, the models in this paper are not described by ordinary differential equation. These models are realized by using the S-function builder of SIMULINK. The steady state calculation result based on the thermal-hydraulic models show agreement with the design value. To show the proposed models could be used for the simulation, the transient process of parameter change is compared with Relap5 simulation code, which shows agreement. A Proportional-Integral (PI) controller is designed to keep the power following the set value as quickly as possible. To keep the inlet temperature of core coolant constant, a control strategy based on a simultaneous feed-forward and feedback scheme has been adopted. The feed-back control system is a PI controller and it can achieve a real time and no error control, but it has time delay. The feed-forward control can realize the control in advance before the LBE temperature at inlet of the core has been changed to reduce the overshoot. So the feed-forward can realize an advance and rough control, the feedback can realize a no error and accurate control. Based on the developed model and control strategy, dynamic simulations of the whole system in case of step changes of reactivity and set power are performed. The simulation results show that the proposed model is accurate enough to describe the dynamic behaviors of the plant in spite of its simplicity. It has also been demonstrated that the developed controllers for the CLEAR-IA can provide superior reactor control due to the efficiency of the control strategy adopted.Copyright