瞬态监督和处理方法在核电厂运行许可证延续中的应用

核电厂延续运行前,由于缺乏瞬态监督管理对核电厂延续寿命影响的具体认知,相关瞬态监督和控制仅限于设计寿期范围内,没有涉及到延续运行。这导致核电厂在运行前期没有针对性地管理瓶颈瞬态的消耗,从而减少了核电厂实际可达的寿命长度;或者相关瞬态数据收集不够详细,不足以支撑更细致的疲劳分析,在延续运行评估时只能采取更多的包络处理,难以实现更长的评估寿命。本文针对上述此问题,通过汲取秦山核电厂延续运行研究中瞬态相关经验,从日常运行监督和专项延续评估两个方面,对核电厂延续运行瞬态监督和数据处理进行研究,形成了适用于核电厂延续运行的瞬态管理技术方法,可有效指导后续核电机组开展延续运行工作。     

百万千瓦级核电厂电网故障甩负荷到厂用电瞬态研究

在核电厂机组运行时，如果母线断路器找开，机组与外电网失去连接，交流发电机继续向机组厂用设备供电，该瞬态称为甩负荷到厂用电，由于电网故障的原因。机组保护系统自动找开母线断路器，称为电网故障甩负荷。甩负荷到厂用电运行可以提高机组的可用性和运行的安全性。本文采用CATIA2－A程序分析了百万千瓦级核电厂在电网故障导致甩负荷到厂用电运行的瞬态变化，由于寿期初、末的反应性系数不同，会导致中子通量峰值的不同，从而决定了瞬态能否成功。     

核电厂汽轮机详细数值建模研究及其瞬态分析

以岭澳一期核电厂汽轮机部件为原型,利用系统程序RELAP5对其进行详细数值建模研究。通过在100%功率稳态工况下的计算证明,详细的汽轮机数值建模弥补了简化建模中焓值计算误差较大的缺陷。将详细的汽轮机数值建模整合到全范围核电厂热力系统模型中进行瞬态分析,并与岭澳一期核电厂原始实验报告中汽轮机负荷从97%功率水平阶跃变化至87%功率水平瞬态运行工况的数据曲线进行对比。结果表明,稳态模型的焓计算值与电厂实际值误差在2%以内,瞬态模型的分析参数趋势符合电厂实际情况。     

压水堆核电厂反应堆轴向功率分布控制

本文介绍了采用低维等效原理和综合法技术,研制成功的压水堆轴向功率偏移控制和负荷跟踪计算程序。该程序为我国秦山二期2×600MW 核电厂反应堆的核设计提供了急需的计算手段。程序的计算精度满足工程设计的要求。本文还利用此程序完成了该反应堆负荷跟踪计算和各种典型的运行功率瞬态分析,初步确定了该反应堆的运行控制方式和轴向功率偏移的安全保护定值。     

模糊神经技术在反应堆安全研究中的应用

概要介绍了模糊神经网络技术 ,并对其在核电厂的负荷跟踪、功率分布控制、运行状况及运行参数的虚拟测量、故障诊断及瞬态识别以及核燃料的质量检查等方面的应用情况进行了综述。模糊神经网络技术在核电厂中的应用大大提高了反应堆运行的安全性和可靠性 ,展现了良好的推广应用前景     

负荷跟踪运行下反应堆一回路控制系统仿真与验证

随着核电厂负荷跟踪运行研究的不断深入,开发针对负荷跟踪过程的仿真模型也势在必行。本文以CNP600压水堆核电厂一回路主系统为研究对象,基于RELAP5/MOD3.4程序建立系统模型,并在此基础上进行控制系统仿真。以典型日负荷跟踪运行模式、负荷线性变化以及负荷阶跃变化等工况瞬态测试对仿真系统进行验证。结果表明,瞬态过程中各参数变化范围和趋势与电厂实际运行值相符,准确反映了负荷跟踪下CNP600压水堆核电厂一回路的运行过程。仿真模型对后续的安全分析具有一定的适用性。     

反应堆轴向功率分布控制和功率能力分析

介绍了采用低维等效原理和综合技术,研制成功的反应堆轴向功率分布控制和功率能力分析的计算程序。程序的计算精度满足工程设计要求。在秦山二期2×600MWe核电厂反应堆核设计中,用该程序完成了该反应堆负荷跟踪计算和各种典型的运行功率瞬态分析,确定了该反应堆的运行控制模式和轴向功率偏移的安全保护定值。     

核电厂主给水泵跳泵导致停堆分析与改进

本文针对核电厂调试试验期间,发生主给水泵意外"跳泵"引发反应堆停堆的根本原因进行技术分析,结合国内同行经验给出了有效的改进措施,从设备本体装配尺寸改进、系统设计回路优化两个方面,对如何提高主给水泵运行安全和可靠性进行深入阐述和分析。经过相关试验运行验证,优化改进措施和方案可以提高设备运行可运行性,尤其是保证主给水泵在负荷变化瞬态时的可运行性。并为其他核电厂同类型主给水泵运行调试,提供了一定的工程参考。     

核电厂控制调节系统在运行瞬态分析中的数值计算

用数值方法模拟核电厂控制调节系统的运行特性,是核电厂运行瞬态分析的重要组成部份。文章介绍了核电厂功率、旁通排放、蒸汽发生器水位等调节系统的数值计算方法;同时对上述调节系统在核电厂事故过程中的安全作用也进行了讨论。     

瞬态曲线分段线性化方法研究及其程序实现

核电厂的瞬态数据是相关设备应力分析和评定的必要输入。然而,电厂实际运行记录数据或设计瞬态计算数据通常由时间间隔很短的大量数据点构成,直接作为应力分析的输入将会导致大量不必要的计算。为了简化分析,一般需要将瞬态数据曲线进行分段线性化处理。本文基于最小二乘法,开发了一种瞬态曲线分段线性化的方法,并且通过编写PLTC(PiecewiseLinearization forTransientCurves)程序予以实现。分别采用标准正弦函数曲线和核电厂典型瞬态曲线对本文的方法进行了验证,结果表明,本文的方法和程序能够很好地实现瞬态曲线的分段线性化。本文的研究有助于提高核电厂瞬态曲线分段线性化的效率和精度。     

Boron transient effects on the behavior of IRIS reactor using dynamic modeling

Small Modular Reactors (SMR) are considered as having several advantages over typical nuclear reactors under various specific conditions. They are thought to be installed in countries with small or medium power grid, in which a large power plant is not necessary or in isolated communities far from distribution centers. A plenty of developing countries are in this situation, so that a significant demand on this type of reactor is expected in a near future. The IRIS reactor is the top-front of SMRs, making its complete development very attractive, since it can fulfill the essential requirements for a future nuclear power plant: better economics, safety-by-design, low proliferation risk and environmental sustainability. IRIS reactor is an integral type PWR in which all primary components are arranged inside the pressure vessel. This configuration involves important changes when compared with a conventional PWR. These changes require several studies to comply with the safe operational limits for the reactor. In light water reactors, a solution of boric acid is used in the coolant of the primary loop to absorb neutrons, aiming to adjust the reactivity of the reactor. A significant decrease in the boron concentration in the core might lead to a considerable power excursion. Several studies on PWR have established correlations between power excursions and deficiencies in homogenization of boric acid diluted in the coolant. The IRIS reactor, due to its integral configuration, does not possess a spray system for boron homogenization which may cause power transients. In this paper, a study has been conducted to develop a dynamic model (named MODIRIS) for transient analysis, implemented in the MATLAB'S software SIMULINK, allowing the analysis of IRIS behavior by considering the neutron point kinetics model for power generation. The methodology is based on generating a set of differential equations of neutronic and thermal-hydraulic balances which describes the dynamics of the primary circuit, as well as a set of differential equations describing the dynamics of secondary circuit. The equations and initialization parameters at full power were inserted into the SIMULINK and the code was validated by comparing with RELAP simulations for a transient of feedwater reduction in the steam generators. Furthermore, the current paper looks for studying and developing a dynamic model for calculating the variations in the boric acid concentration. Then, a simplified model for boron dispersion was implemented into the code MODIRIS to simulate power transients which occur due to variations in the boron concentration in the primary loop of the IRIS reactor. The results for boron concentration, inserted reactivity and steam production showed a good precision and represented the expected behavior very well in the range of operational transients.     

1GW固态燃料熔盐堆运行瞬态分析

钍基熔盐堆(Thorium-based Molten Salt Reactor,TMSR)作为一种新的堆型,具有独特的安全与运行特性。研究其热工水力特性,对其进行瞬态分析,将有助于深刻理解该反应堆。本文介绍了1 GW固态熔盐堆的堆芯设计方案,并描述了用于瞬态分析的详细程序结构。其中,利用RELAP5对其热工水力模型进行模拟;利用Simulink对其控制系统模型进行模拟。通过预期运行瞬态,例如功率降低、堆芯反应性引入、二回路温度变化等工况显示了其运行特性,并验证了控制系统可以使反应堆达到安全稳定状态,而不触发保护系统动作。     

Comparison of two models for a high temperature reactor coupled to a gas turbine

This paper presents two independent dynamic models of a nuclear gas turbine power plant. Both the high temperature nuclear reactor (HTR) and its energy conversion system (ECS) based on a direct Brayton cycle have been modelled. One model utilises RELAP5 for the ECS, the other Aspen Custom Modeler (ACM). The reactor model used in both models is a point kinetic model derived from a detailed reactor model. The ECS model is described and compared componentwise, with an emphasis on the turbomachinery. The total plant models are compared with each other by calculating two representative transients: one load rejection transient and one transient with the system at part load.     

船用核动力装置净化系统运行安全分析研究

基于船用核动力装置运行安全的需要,阐述运行安全分析研究的特点和重要性。开发了核动力装置全范围运行分析软件,并对一回路净化系统典型事故进行了运行分析,给出了事故判别和处置的措施。所用运行分析软件和方法可满足船用核动力装置动态特性分析、事故处理规程制定和验证等工作的需要。     

Identification of nuclear power plant transients using the Particle Swarm Optimization algorithm

In order to help nuclear power plant operator reduce his cognitive load and increase his available time to maintain the plant operating in a safe condition, transient identification systems have been devised to help operators identify possible plant transients and take fast and right corrective actions in due time. In the design of classification systems for identification of nuclear power plants transients, several artificial intelligence techniques, involving expert systems, neuro-fuzzy and genetic algorithms have been used. In this work we explore the ability of the Particle Swarm Optimization algorithm (PSO) as a tool for optimizing a distance-based discrimination transient classification method, giving also an innovative solution for searching the best set of prototypes for identification of transients. The Particle Swarm Optimization algorithm was successfully applied to the optimization of a nuclear power plant transient identification problem. Comparing the PSO to similar methods found in literature it has shown better results.     

Experimental study of natural circulation instability with void reactivity feedback during startup transients for a BWR-type SMR

The natural circulation boiling type SMR can experience flow instability during the startup transients due to the void reactivity feedback. A BWR-type natural circulation test loop has been built to perform the nuclear coupled startup transient tests for Purdue Novel Modular Reactor (NMR). This test loop is installed with different instruments to measure various thermal hydraulic parameters. The testing process can be monitored and controlled through PC with the assistance of LabVIEW procedure. The effects of power ramp rate on the flow instability during the nuclear coupled tests were investigated by controlling the power supply based on the point kinetics model with coolant void reactivity feedback. Two power ramp rates were investigated and the results were compared with those of the thermal hydraulic startup transients without void reactivity feedback. The time trace of power supply, system pressure, natural circulation rate, and void fraction profile are used to determine the flow stability during the transients. The results show that nuclear coupled startup transients also experience flashing instability and density wave oscillations. The power curves calculated from point kinetics model for startup transients show some fluctuations due to void reactivity feedback. However, the void reactivity feedback does not have significant effects on the flow instability during the startup procedure for the NMR.     

Analysis of the Leibstadt power plant condensate and feedwater systems during selected operational transients

This paper presents a selection of plant analyses that were carried out by PSI in support of the Leibstadt Nuclear Power Plant (Swiss boiling water reactor). The analyses were performed as part of a collaboration between Leibstadt and PSI, to help resolve some operational problems that were experienced during the power uprate beginning in 1998. The issues under investigation were related to the behavior of the condensate and feedwater systems during transients initiated by a turbine trip, load rejection and a single feedwater pump trip, all of which increased the risk of an inadvertent reactor shutdown by reaching reactor pressure vessel water level limits. The possibility of a reactor shutdown was related to perturbations in the feedwater flow caused by transitory pump cavitation of the feedwater pumps, due to a rapid depressurization in the feedwater tank. In addition to a direct analysis of plant measurement provided by Leibstadt, steady-state and transient simulations of the events were performed at PSI using the system codes TRAC-BF1 and TRACE. Through a combination of the analysis of the plant measurements, the code simulations and an analysis of the whole plant behavior using the Leibstadt plant-simulator appropriate modifications of the plant hardware, control system and operational set points were proposed. The implementation and success of these changes were verified by a number of plant tests. Finally, the original designed plant capability not to shutdown during the aforementioned transients was demonstrated.     

Mode-C运行与控制模式设计技术研究

本文基于运行与控制模式设计,结合核电厂的运行需求,针对国内压水堆核电厂以基负荷运行方式为主、负荷跟踪运行需求较少的特点,首次开展了与之适应的Mode-C运行与控制模式设计。通过控制策略设计、控制棒设置设计、核电厂运行方式设计、核电厂运行范围设计等设计步骤,研究Mode-C运行与控制模式的设计技术。结果表明：采用Mode-C模式的压水堆核电厂能根据负荷变化需求选择执行单变量自动控制模式或双变量自动控制模式,实现了设定的控制策略,Mode-C运行与控制模式的设计技术在反应堆物理专业方面是可行的。     

Transient flow analysis in reactor coolant pump systems during flow coastdown period

The rapid flow transient calculation in reactor coolant pump system is important in the safety analysis of a nuclear reactor. An accurate transient analysis of flow coastdown is also important and necessary for the design and manufacture of a reactor coolant pump. Only under the reliable work of a reactor coolant pump the safety of a nuclear power plant can be guaranteed. A mathematical model is developed for solving flow rate transient and pump speed transient during flow coastdown period. The detailed information of the centrifugal pump characteristics is not required. The flow rate and pump speed are solved analytically. The analytic solution of non-dimensional flow rate indicates that non-dimensional flow rate is determined by energy ratio β. The kinetic energy of the loop coolant fluid and the kinetic energy stored in the rotating parts are two important parameters in form of β. When the steady-state flow rate and pump speed are constant, the inertia of primary loop fluid and the pump moment of inertia are also two important parameters in flow transient analysis. For the condition all pump shafts are seized, the flow decay depends on the inertia of primary loop fluid. For the case that pump inertia is very large, the flow decay is determined by the pump inertia. The calculated non-dimensional flow rate and non-dimensional pump speed using the model are compared with published experimental data of two nuclear power plants and a reactor model test on flow coastdown transients. The comparison results show a good agreement. As the flow rate approaches to zero, the increase difference between experimental and calculated value is due to the effect of the mechanical friction loss.     

A system analysis tool with a 2D gas turbine modeling for the load transients of HTGRS

This paper presents the operational performance and transient response of a high temperature gas-cooled reactor (HTGR) with an emphasis on the gas turbine through a two-dimensional approach. For its operational and transient simulation we use a GAMMA-T in which the system code, GAMMA, is coupled with the two-dimensional turbomachinery model. We also implement several models into the GAMMA-T: the reactor kinetics model, the bypass valve model, and the models of the core, the heat exchangers, the gas turbine, and the piping. The estimations of compressor and turbine performances are based on a two-dimensional axisymmetric throughflow method that is capable of predicting both the transient and steady-state behavior of the power conversion system (PCS). To demonstrate the code capability, we investigated the two representative transients of GTHTR300, which is a 600 MW direct cycle helium cooled reactor consisting of a prismatic block type core, a horizontal single-shaft configuration of turbomachinery, a recuperator, and a precooler: a loss of heat rejection transient corresponding to the failure of the precooler water supply, and a 30% load reduction transient from nominal operation with bypass control. The simulation results demonstrated the controllability and operational stability for the plant.