基于向量通用发生函数理论的考虑多性能参数的热力系统可用度分析

研究了将向量通用发生函数理论用于对考虑多性能参数的复杂热力系统进行可用度分析。定义了设备性能的向量通用发生函数,建立了热力系统可用度分析的向量发生函数算法模型,给出了多状态设备状态概率的计算方法。通过对系统内部多因素退化规律的随机模拟,对复合因素影响下的多性能参数热力系统可用度进行了分析计算。实例研究表明,传统的基于二态理论的系统可用度评估结果偏保守,而基于多状态系统理论考虑参数失效的系统可用度评估结果更能反映系统的实际使用特性。     

标准M310高压安注系统可靠性分析

本文分析标准 M310 核电站高压安注系统的设计特点,用分期系统可靠性分析方法计算该系统的不可用度,并给出定量计算结果和分析结论。     

基于使用可用度的核岛外设备机会维修策略研究

针对核电站事故停机将导致较大经济损失且核岛外设备允许实施并行维修的特点,对核岛外设备机会维修策略进行了研究,提出了一种以设备自身最优维修间隔期为基准的机会维修控制策略.利用蒙特卡罗仿真方法,给出了该策略中以使用可用度最大为目标的最优维修问隔期和机会维修系数的仿真求解算法,并用算例证明了采取机会维修策略能有效提高系统的使用可用度.     

中国实验快堆380V应急交流电力系统的可靠性分析

针对中国实验快堆380 V应急交流电力系统的系统设计及功能,运用经典故障树分析方法对其进行可靠性评价。定性分析得到了导致系统失效的支配性最小割集,定量计算得到系统的不可用度为6.65E-12,以及最小割集重要度和灵敏度,从而对该系统的维修和试验提供参考。     

考虑性能退化和多级保障的核动力系统可用度分析

研究了将通用发生函数理论与Markov过程方法相结合的方法对考虑性能退化和多级保障的核动力系统进行可用度分析。基于Markov过程方法建立考虑性能退化和多级保障的设备状态概率模型,并将其嵌入由发生函数构建的系统逻辑关系模型和性能状态组合模型,从而分析系统在不同性能需求条件下的可用度,并分析不同修理条件对系统可用度的影响。所建立的模型反映了核动力系统的使用特点和维修保障特性,研究方法和分析结论能为核动力系统的使用管理、维修决策及保障条件分析提供理论指导和依据。     

基于UGF和Semi-Markov方法的反应堆泵机组多状态可靠性分析

将通用发生函数(UGF)与半马尔可夫过程(Semi-Markov Process)相结合,对反应堆泵机组进行多状态可靠性分析。给出多状态系统可靠性分析的UGF算法模型,推导多状态设备性能状态的Semi-Markov过程概率表达式,定义设备、系统性能值的定量描述方法。以性能参数是否满足需求值做为系统成功与失效的判据,对比分析反应堆泵机组在需求性能条件下的多状态可用度与2状态可用度结果,并给出系统在任务周期内的平均性能值。结果表明,该方法能够定量分析部分失效对系统可靠性的影响,降低传统的2态可靠性分析方法产生的不必要的保守程度。     

非安全级DCS预期瞬态不停堆系统可靠性分析

介绍了非安全级DCS平台的结构,将预期瞬态不停堆系统按功能划分,每个功能由输入模块、控制器模块、输出模块组成。研究了各种模块组合结构的可靠性参数及其串联组成的系统可靠性。针对多模块组合结构的马尔可夫模型过于复杂,计算量大的问题,提出了优化的建模方法。基于优化的马尔可夫模型,计算了3种数字量输入模块组合结构的拒动率和误动率指标,得出相同模块类型和数量的情况下,有软件表决逻辑的2种组合结构可靠性指标较高,而由硬件实现的组合结构可靠性指标较低的结论,为组合结构的选择提供了设计依据。     

核反应堆专设安全设施试验间隔期的确定

针对核反应堆专设安全设施试验间隔期的优化,以传统的专设安全设施组成设备平均可用度模型为基础,提出采用较少假设近似并考虑设备在备用、试验和检修3阶段状态关联的可用度模型。改进模型比原模型具有更为广泛的适用性,对大型复杂系统、精度要求比较高的分析计算具有优势。将该模型用于核反应堆余热排出系统试验间隔期的确定,结果能够为核反应堆专设安全设施的使用管理和维修决策提供参考。     

核反应堆专设安全设施试验间隔期的确定

针对核反应堆专设安全设施试验间隔期的优化,以传统的专设安全设施组成设备平均可用度模型为基础,提出采用较少假设近似并考虑设备在备用、试验和检修3阶段状态关联的可用度模型。改进模型比原模型具有更为广泛的适用性,对大型复杂系统、精度要求比较高的分析计算具有优势。将该模型用于核反应堆余热排出系统试验间隔期的确定,结果能够为核反应堆专设安全设施的使用管理和维修决策提供参考。     

印制电路板X射线瞬态响应计算

印制电路板的X射线瞬态辐照响应是系统电磁脉冲(SGEMP)效应主要研究内容之一。脉冲X射线辐照后,电荷的重新分布在印制电路板布线上会引起电压电流响应。建立了两种结构印制电路板瞬态辐照响应的传输线计算模型,并介绍了计算模型中电流源、电容参数的物理意义及其计算方法;该模型考虑了电子的运动规律及响应的产生机理,适用于瞬态辐照响应规律研究和工程预估,采用的方法和思想可以应用于其它复杂结构印制电路板瞬态辐照响应的模拟。最后,给出了一个计算实例。     

Development and assessment of fire-related risk unavailability matrices to support the application of the maintenance rule in a PWR nuclear power plant

Two methods are presented which serve to incorporate the fire-related risk into the current practices in nuclear power plants with respect to the assessment of configurations. The development of these methods is restricted to the compulsory use of fire probabilistic safety assessment (PSA) models. The first method is a fire protection systems and key safety functions unavailability matrix which is developed to identify structures, systems, and components significant for fire-related risk. The second method is a fire zones and key safety functions (KSFs) fire risk matrix which is useful to identify fire zones which are candidates for risk management actions. Specific selection and quantification methodologies have been developed to obtain the matrices. The Monte Carlo method has been used to assess the uncertainty of the unavailability matrix. The analysis shows that the uncertainty is sufficiently bounded. The significant fire-related risk is localized in six KSF representative components and one fire protection system which should be included in the maintenance rule. The unavailability of fire protection systems does not significantly affect the risk. The fire risk matrix identifies the fire zones that contribute the most to the fire-related risk. These zones belong to the control building and electric penetrations building.     

Mechanical System Surveillance Test Interval Optimization

Abstract

In general, functional tests are required periodically to detect failures in standby equipment as a means of assuring their availability on demand. However, these functional tests may have an adverse impact on the system availability, because of their potential negative effects such as unavailability at the period of testing, increase of failure rate resulting from excessive wear due to testing. Therefore, in principle, there is an optimal test interval for each standby system. As it normally is difficult to determine analytically the optimal test interval for the real systems in the industry, we therefore introduce the Monte Carlo simulation method for this purpose. In this paper, a one-component system was chosen firstly to introduce the concept of test interval optimization as well as the feasibility of using Monte Carlo simulation. Secondly, a real system in the nuclear power plant. Standby-liquid-control system, was studied using Monte Carlo simulation technique. The promising results obtained confirm the concept and methodology used.     

计算高可靠性系统失效概率的统计估计蒙特卡罗方法

在相似仿真方法的基础上 ,设计了计算系统失效概率的统计估计蒙特卡罗方法 ,包括直接统计估计和加权统计估计蒙特卡罗方法。介绍了统计估计蒙特卡罗可靠性仿真的基本原理 ,给出了统计估计蒙特卡罗计算方法的无偏估计量和具体算法。同时采用直接仿真方法、限制抽样蒙特卡罗方法、强迫转换蒙特卡罗方法、直接统计估计和加权统计估计蒙特卡罗方法计算了一高可靠性系统的失效概率 ,结果表明 ,在高可靠性系统不可靠度计算中加权统计估计蒙特卡罗方法计算结果的方差最小 ,效率最高。     

Monte Carlo Calculations of Irradiation Test Photon Spectra

Calculations of scattered photon spectra in concrete rooms are made, by means of the Monte Carlo method, for Co-60 sources of the type used in irradiation testing of electronic devices. It is found that the scattered photon spectrum shape is heavily dependent on the location of source and target, and that in devices where dose enhancement can occur, there can be substantial variation in the absorbed dose due to this strong dependence on location. Scattered photon spectra are also obtained for two types of irradiation test cells. Additionally, dose enhancement ratios obtained with a coupled photon-electron Monte Carlo calculation are given for a gold-metallized silicon device.     

Probabilistic simulation of cable performance and water based protection in cable tunnel fires

Nuclear power plants contain a significant amount of fire load in form of electrical cables. The performance of the cables is interesting both from the fire development and system failure viewpoints. In this work, cable tunnel fires are studied using numerical simulations, focusing on the fire spreading along power cables and the efficiency of the water suppression in preventing the cable failures. Probabilistic simulations are performed using Monte Carlo technique and the Fire Dynamics Simulator (FDS) as the deterministic fire model. The primary fire load, i.e. the power cables are modelled using the one-dimensional pyrolysis model, for which the material parameters are estimated from the experimental data. Two different water suppression systems are studied. The simulation results indicate that using either suppression system decreased the heat release rate in the tunnel to less than 10% of the non-suppressed case. Without water suppression, the cables of the second sub-system were damaged in almost all fires, but when either of the studied water suppression systems was used, the probability of the cable failures was decreased to less than 1%. This result indicates that in current scenario, the probability of losing both sub-systems is determined directly by the suppression system unavailability.     

A Proposal on Data Treatment In the Feynman-α Experiment

A new algorithm of Monte Carlo criticality calculations for implementing Wielandt's method, which is one of acceleration techniques for deterministic source iteration methods, is developed, and the algorithm can be successfully implemented into MCNP code. In this algorithm, part of fission neutrons emitted during random walk processes are tracked within the current cycle, and thus a fission source distribution used in the next cycle spread more widely. Applying this method intensifies a neutron interaction effect even in a loosely-coupled array where conventional Monte Carlo criticality calculation methods have difficulties, and a converged fission source distribution can be obtained with fewer cycles. Computing time spent for one cycle, however, increases because of tracking fission neutrons within the current cycle, which eventually results in an increase of total computing time up to convergence. In addition, statistical fluctuations of a fission source distribution in a cycle are worsened by applying Wielandt's method to Monte Carlo criticality calculations. However, since a fission source convergence is attained with fewer source iterations, a reliable determination of convergence can easily be made even in a system with a slow convergence. This acceleration method is expected to contribute to prevention of incorrect Monte Carlo criticality calculations.     

Development of a practical Monte Carlo based fuel management system for the Penn State University Breazeale Research Reactor (PSBR)

A practical fuel management system for the he Pennsylvania State University Breazeale Research Reactor (PSBR) based on the advanced Monte Carlo methodology was developed from the existing fuel management tool in this research. Several modeling improvements were implemented to the old system. The improved fuel management system can now utilize the burnup dependent cross section libraries generated specifically for PSBR fuel and it is also able to update the cross sections of these libraries by the Monte Carlo calculation automatically. Considerations were given to balance the computation time and the accuracy of the cross section update. Thus, certain types of a limited number of isotopes, which are considered “important”, are calculated and updated by the scheme. Moreover, the depletion algorithm of the existing fuel management tool was replaced from the predictor only to the predictor-corrector depletion scheme to account for burnup spectrum changes during the burnup step more accurately. An intermediate verification of the fuel management system was performed to assess the correctness of the newly implemented schemes against HELIOS. It was found that the agreement of both codes is good when the same energy released per fission (Q values) is used. Furthermore, to be able to model the reactor at various temperatures, the fuel management tool is able to utilize automatically the continuous cross sections generated at different temperatures. Other additional useful capabilities were also added to the fuel management tool to make it easy to use and be practical. As part of the development, a hybrid nodal diffusion/Monte Carlo calculation was devised to speed up the Monte Carlo calculation by providing more converged initial source distribution for the Monte Carlo calculation from the nodal diffusion calculation. Finally, the fuel management system was validated against the measured data using several actual PSBR core loadings. The agreement of the predicted core excess reactivities and the measured values is found to be good considering the measurement uncertainties.     

Monte Carlo methods for radiation transport analysis on vector computers

The development of advanced computers with special capabilities for vectorized or parallel calculations demands the development of new calculational methods. The very nature of the Monte Carlo process precludes direct conversion of old (scalar) codes to the new machines. Instead, major changes in global algorithms and careful selection of compatible physics treatments are required. Recent results for Monte Carlo in multigroup shielding applications and in continuous-energy reactor lattice analysis have demonstrated that Monte Carlo methods can be successfully vectorized. The significant effort required for stylized coding and major algorithmic changes is worthwhile, and significant gains in computational efficiency are realized. Speedups of at least twenty to forty times faster than CDC-7600 scalar calculations have been achieved on the CYBER-205 without sacrificing the accuracy of standard Monte Carlo methods. Speedups of this magnitude provide reductions in statistical uncertainties for a given amount of computing time, permit more detailed and realistic problems to be analyzed, and make the Monte Carlo method more accessible to nuclear analysts. Following overviews of the Monte Carlo method for particle transport analysis and of vector computer hardware and software characteristics, both general and specific aspects of the vectorization of Monte Carlo are discussed. Finally, numerical results obtained from vectorized Monte Carlo codes run on the CYBER-205 are presented.     

中子-光子输运程序TOPAN的开发与验证

开发的粒子输运蒙特卡罗通用程序TOPAN能模拟中子和光子的耦合输运,可给出中子和光子在介质中输运后的点通量、面通量、体通量、面流量等参数。该程序除具备常用的蒙特卡罗软件功能外,还增加了处理介质温度变化的等效质量热运动模型和非均匀介质中粒子输运模块,具备粒子标识功能,初步具备了进行一些复杂问题中粒子输运模拟的能力。结合具体算例对TOPAN程序的各功能模块进行了比对验证。     

三维离散纵标-蒙特卡罗耦合系统TDOMINO开发与验证

基于离散纵标和蒙特卡罗方法开发了三维离散纵标 蒙特卡罗耦合系统TDOMINO,其耦合形式灵活,可根据需要选择不同坐标系下的耦合方式进行计算分析。利用美国橡树岭国家实验室提供的HBR-2基准题,采用TDOMINO分别建立了直角坐标系和圆柱坐标系下的耦合模型进行验证计算,给出了反应堆辐照监督管处6种典型核素比活度的计算结果,与基准报告中提供的实验测量值和DORT、MCNP、TORT等程序计算结果相比,TDOMINO具有较好的计算精度,可用于解决复杂屏蔽计算问题。