A New Method for Low Sensitivity Design of a Controller for a Load Following BWR Power Plant

A new method has been proposed in which the controller is approximated by two low order models, the second being the sensitivity model of the first. Thus a Kind of piecewise solution is proposed in which the system retains the same order irrespective of the number of parameters to be considered for low sensitivity design. The usefulness of the proposed technique is illustrated in the controller design for a direct cycle BWR power plant of 457 MW(thermal) with recirculation control. The mathematical model includes the reactor kinetics, hydrodynamics of the recirculation loop, pressure transients, and the load frequency control system. The response of system variables such as frequency, neutron power, and reactor pressure are plotted with the low sensitivitty controller. The sensitivity function of frequency has been plotted using the conventional and proposed low sensitivity controller for 20 per cent variation in parameter values. The method is specifically recommended for controller design of large size systems.     

Reactivity Control Method for Extended Burnup of Boiling Water Reactor Fuel Bundles

Abstract

A reactivity control method was proposed for a boiling water reactor (BWR) fuel bundle, which has a potential for higher burnup with an increase in fuel enrichment. The new method optimized the distribution and amount of nonboiling water area in a fuel bundle in order to enhance the reactivity control capacity.

Using the method, a 9×9 lattice fuel bundle with a small-sized channel box, large-sized water rods and a reduced fuel rod diameter was proposed for the discharged burnup of 70 GWd/t and the operational cycle length of 18 months. The core, which consists of the proposed fuel bundles with the bundle-averaged enrichment of 5.8% and includes other modifications concerning a neutron low leakage loading pattern, natural uranium axial blankets, and spectral shift with recirculation flow control, has a cold shutdown margin greater than the design limit (1%Δk) with minimum fuel bundle shuffling. Further, it has potentials for natural uranium savings of about 20% per unit power and reduction in the amount of reprocessing of about 60% per unit power, compared with current BWR designs.     

Simplified distributed parameters BWR dynamic model for transient and stability analysis

This paper describes a simplified model to perform transient and linear stability analysis for a typical boiling water reactor (BWR). The simplified transient model was based in lumped and distributed parameters approximations, which includes vessel dome and the downcomer, recirculation loops, neutron process, fuel pin temperature distribution, lower and upper plenums reactor core and pressure and level controls. The stability was determined by studying the linearized versions of the equations representing the BWR system in the frequency domain. Numerical examples are used to illustrate the wide application of the simplified BWR model. We concluded that this simplified model describes properly the dynamic of a BWR and can be used for safety analysis or as a first approach in the design of an advanced BWR.     

蒸汽发生器水位双PI控制的改进研究

蒸汽发生器（SG）水位的控制问题是压水堆核电站紧急停堆的1个重要原因,尤其在低功率情况下,缩胀效应致使水位控制较难。为了避开水位变化初期的假水位的影响,在控制过程中加入判别器来决定水位信号是否参与控制。采用遗传算法对控制器参数进行整定,采用的SG水位模型是简化的数学模型。对控制器中有无判别器的控制效果进行了对比。对比结果表明：加入判别器后,缩短了调节时间,降低了超调量。     

快堆冷却剂出口温度的自抗扰控制

为提高快堆控制系统的瞬态响应速度,降低控制器的超调量及稳态误差,设计了快堆冷却剂出口温度的线性自抗扰控制器。基于快堆的中子动力学方程、反应性方程和堆芯热传输方程,推导出了用于线性自抗扰控制器设计的带总扰动项的二阶模型。利用所得二阶模型的参数,确定了带模型信息的线性扩展状态观测器的部分参数。最后,在MATLAB环境下对控制器模型进行了仿真并加入系统模型信息。仿真结果表明,所设计的控制器具有良好的性能。相比于以冷却剂流量变化量为控制量的控制器,以冷却剂流量作为控制量的控制器具有更快的响应速度、更小的超调量、更好的抗干扰能力。     

Effect of automatic recirculation flow control on the transient response for Lungmen ABWR plant

In this study the automatic mode of the recirculation flow control system (RFCS) for the Lungmen ABWR plant has been modeled and incorporated into the basic RETRAN-02 system model. The integrated system model is then used to perform the analyses for the two transients in which the automatic RFCS is involved. The two transients selected are: (1) one reactor internal pump (RIP) trip, and (2) loss of feedwater heating. In general, the integrated system model can predict well the response of key system parameters, including neutron flux, steam dome pressure, heat flux, RIP flow, core inlet flow, feedwater flow, steam flow, and reactor water level. The transients are also analyzed for manual RFCS case, between the automatic RFCS and the manual RFCS cases, comparisons of the transient response for the key system parameter show that the difference of transient response can be clearly identified. Also, the results show that the ΔCPR (delta critical power ratio) for the transients analyzed may not be less limiting for the automatic RFCS case under certain combination of control system settings.     

Evaluation of Thermal Margin during BWR Neutron Flux Oscillation

Fuel integrity is very important, from the view point of nuclear power plant safety. Recently, neutron flux oscillations were observed at several BWR plants. The present paper describes the evaluations of the thermal margin during BWR neutron flux oscillations, using a three- dimensional transient code. The thermal margin is evaluated as MCPR (minimum critical power ratio).

The LaSalle-2 event was simulated and the MCPR during the event was evaluated. It was a core-wide oscillation, at which a large neutron flux oscillation amplitude was observed. The results indicate that the MCPR had a sufficient margin with regard to the design limit.

A regional oscillation mode, which is different from a core-wide oscillation, was simulated and the MCPR response was compared with that for the LaSalle-2 event. The MCPR decrement is greater in the regional oscillation, than in the core wide-oscillation, because of the sensitivity difference in a flow-to-power gain.

A study was carried out about regional oscillation detectability, from the MCPR response view point. Even in a hypothetically severe case, the regional oscillation is detectable by LPRM signals.     

T-S型模糊切换控制器在堆芯功率控制中的应用#br# #br#

The traditional PID controller is used to control the core power, which has the problems of large overshoot and long regulating time in the control process. In order to solve this problem, based on the core transfer function model, the PD controller, the PID controller and the fuzzy controller are weighted and switched by T-S fuzzy rules, and T-S fuzzy switching controller is designed. Taking the core power control of a lead cooled fast reactor as an example, a T-S fuzzy switching control system of the core power is established to simulate the relative power setpoint value step and the core inlet coolant temperature disturbance. The results show that the T-S fuzzy switching controller designed based on the core transfer function model can achieve a good control of the core power.     

Semi-Time-Optimal Control of Boiling Water Reactor

A simple semi-time-optimal controller is proposed for a linear single-input single-output system. The controller is presented as a state feedback with time-varying gains. Stability analysis of the control system shows that the system is stable if the closed loop coefficient matrix is within a limited range. A method of eigenvalue assignment for the closed loop is provided and coordinate transformation is applied in the present paper to design the controller. Simulations of the control system for BWR indicate that the present controller realizes the time-optimal control better than the conventional PID controller.     

An investigation of the dynamics of nuclear power facilities upon deterioration of heat exchange

Conclusions The structure of a control system with independent setting of the neutron power and the revolution rate of the pump of the second loop directly from a controller without additional measures for stabilization is unstable in facilities with a once-through steam generator.A similar structure in facilities in which a steam generator with recirculation is used is stable under normal operating conditions and gives a satisfactory quality of the transitional processes. However, one should bear in mind that such a structure is unstable in facilities in which thermal loads are close to the critical ones when a heat-exchange crisis arises in the steam generator. In order to eliminate the cited instability, it is necessary to introduce relations into the structure of the control system which form a law for controlling the reactor power as a function of directly measurable variables which depend on the temperature conditions in the reactor.Translated from Atomnaya Énergiya Vol. 46, No. 5, pp. 320–324, May, 1979.     

A semiempirical prediction of the decay ratio for the boiling water reactors start-up process

During the start-up of a commercial boiling water reactor (BWR), the power and the coolant flow are continuously monitored. In order to prevent power instability events, the decay ratio (DR) could also be monitored. The process can be made safer if the operator could anticipate the DR too. DR depends on the power, the flow and many other quantities such as axial and radial neutron flux distribution, feed water temperature, void fraction, etc. A simple relationship for DR is derived. Three independent variables seem to be enough: the power, the flow and a single parameter standing for all other quantities which affect the DR. The relationship is validated with data from commercial BWR start-ups. A practical procedure for the start-up of a BWR is designed; it could help preventing instability events.     

基于变论域模糊PID的液态熔盐堆堆芯功率控制

液态熔盐堆堆芯系统具有非线性、时变性等特点,模糊比例积分微分（PID）控制技术因初始论域不能跟随误差变化而伸缩,使得系统的控制精度降低,故设计了一种基于变论域模糊PID控制器的堆芯功率控制策略。以熔盐增殖堆MSBR堆芯为例,在堆芯入口温度扰动或堆芯反应性扰动下,使用Matlab/Simulink对PID控制、模糊PID控制与变论域模糊PID控制下的效果进行仿真对比。结果表明,基于变论域模糊PID控制器建立的堆芯功率控制系统响应速度更快,超调量更小,控制效果更佳。      

协调控制技术在核动力系统中的应用研究

在传统PID控制的基础上引入了协调控制器,以此组成核动力装置的协调控制系统.在建立核动力系统模型的基础上应用协调控制系统并进行了仿真试验,仿真结果表明:相比于传统PID控制,采用协调控制后,冷却剂平均温度和出口蒸汽压力的超调量减小,稳定时间变短,而汽轮机功率响应稍微慢了一点,但其峰值超调变小了.     

Application of a self-organizing fuzzy logic controller to nuclear steam generator level control

In this paper, the self-organizing fuzzy logic controller is investigated for the water level control of a steam generator. In comparison with conventional fuzzy logic controllers, this controller performs the control task with no initial control rules; instead, it creates control rules and tunes input membership functions based on the performance criteria as the control behavior develops, and also modifies its control structure when uncertain disturbance is suspected. Selected tuning parameters of the self-organizing fuzzy logic controller are updated on-line in the learning algorithm, by a gradient descent method. This control algorithm is applied to the water level control of a steam generator model developed by Irving et al. The computer simulation results confirm the good performance of this control algorithm for all power ranges. This control algorithm can be expected to be used for the automatic control of a feedwater control system in a nuclear power plant with digital instrumentation and control systems.     

基于LMIs的核电站蒸汽发生器水位控制策略研究

核电站蒸汽发生器水位常因多种干扰而具有不确定性,因此传统线性水位控制系统较难满足核电站运行要求。针对具有一定不确定性的核电站蒸汽发生器水位模型,根据其特性,借助线性矩阵不等式(LMIs)设计状态反馈H^∞控制器、基于极点配置的H^∞控制器以及H^∞跟踪控制器。在不同稳定工作点下通过仿真试验对上述控制器的效果进行验证和比较,得出了不同稳定工作点下控制策略的特性。     

秦山第二核电厂蒸汽发生器液位控制系统PID参数整定

基于核电厂SimPort仿真平台构建了秦山第二核电厂I号机组蒸汽发生器液位控制系统仿真模型,在该仿真模型上进行了各种工况下的瞬态仿真实验和研究,获得了液位控制系统PID参数的整定值,其中液位控制器的Kp=4．25,T1=425s,TD=10s;流量控制器的KP=1．0,T1=13s。这些参数整定值与实际值基本一致,可供工程技术人员参考。     

A Method for the Burnup Analysis of Power Reactors in Equilibrium Operation Cycles

This paper describes a method based on the consistent thermal neutron flux concept for burnup analysis of power reactors in equilibrium operation cycle. Radial flux distributions are discussed, as well as the variation of isotopic densities in the fuel, and considerations are given on control rod distribution for maintaining a consistent thermal flux distribution. The present method can be applied to most power reactors like BWR and PWR in which a multiple-batch refueling scheme is adopted.     

Hydraulic study on recirculation loops using computational fluid dynamics (CFD). Design optimization and turbulence reduction

A full scope study of the recirculation system flow of a BWR6 nuclear plant is essential, to completing the characterisation and understanding the dynamics of the flow and specially the bistable flow in the recirculation loop.The finite element model serves to study the sensitivity of the design and to reduce turbulence at the connection between the recirculation pump impulsion piping and the manifold. It is found that the problem lies in the poor hydraulic design of the manifold. As a result of this analysis, in order to reduce turbulence and balance flows in the jet pumps the manifold diameter needs to be increased. The simulation shows the unbalanced flow to the jet pump pipes. This fact can be checked in any nuclear power plant BWR.     

基于NSGA-Ⅱ算法的核反应堆功率LQG控制器设计

反应堆功率控制系统的设计与核电厂的安全性和经济性息息相关。为提高其功率控制性能,本研究以某压水堆核电厂为研究对象,建立了其非线性动态数学模型,推导了其状态空间模型;采用线性二次型高斯（LQG）最优控制方法,设计了堆芯功率控制器;进一步基于遗传算法NSGA-Ⅱ对LQR权重系数进行了多目标优化;将本文所设计的控制器与传统PID控制器进行了典型工况的仿真对比。结果表明,与传统PID控制器相比,基于NSGA-Ⅱ方法优化的LQG控制器不但响应快速、控制精度高、稳定性好,而且具有良好的鲁棒性,能获得更优越的堆芯功率控制效果。     

On the use of the multivariate autoregressive and relative power contribution models for determination of transient root cause

The multivariate autoregressive (MAR) model and the relative power contribution (RPC) ratio are used in this work to determine the root causes of a power oscillation event and an apparent positive reactivity insertion transient occurred at the at the BWR/5 Units of the Laguna Verde Nuclear Power Plant (LVNPP) of México. The application of the MAR and PRC models leads to identify dominant frequencies and the contribution from other different signals to the dominant frequencies. The methodology firstly uses a linear model to estimate the response characteristics of the system and the spectra of the noise sources. The estimate of the process linear predictor is obtained by the ordinary least squares method. Then, the model performs a MAR analysis, and the RPC ratio is computed to determine the inter-relationship between the different reactor noise signals. The RPC ratio is an indication of how the fluctuation of one variable depends on other variables, at each frequency.Reactor signals acquired during the two transient operational events are used in the analysis. In the first event, a problem on the position controller of the flow control valve's stem induced the appearance of a power peak of 12% amplitude on the average power range monitors. Actual insertion of positive reactivity did not occur. The signals used for the analysis come from an average power range monitor, the position of the stem in the valve, controller of stem position, and controller of the recirculation flow. For the second transient, power oscillations of about 12% amplitude occurred. Signals from an average power range monitor, total flow through the core and flow through the 10 jet pump of each loop are analyzed. In both cases, some other signals were also used, but since they did not show appreciable influence on the RPC results, they were not considered for final analysis.The RPC results obtained in here confirmed previously known facts about the origin of the transients analyzed. Specifically, for the first transient, a dominant frequency of 1.7 Hz appeared on the power spectral densities of different signals from sensors on the recirculation loop B. At this 1.7 Hz frequency, the RPC ratio showed influence of such loop B spectra to the average reactor power spectrum, but no influence at all of the average reactor power to any of such loop B signals. The root cause, although not a real power transient event, therefore was not of neutronic nature, but related to recirculation flow. In the second transient, the prominent power oscillation frequency (0.54 Hz) was tracked within the spectral data of other signals. The RPC results for this case showed a strong influence of the average reactor power on the flow signals, but only a modest contribution from the flow signals to the average reactor power. The root cause therefore was of neutronic nature, due mainly to a combination of a particular core configuration and control rod pattern change at the moment of the event.