基于模糊综合评价的低压安全注射系统可靠性分析

结合设备运行过程中的专家意见,通过模糊综合评价的方法,对专家评价数据进行分析,得到故障模式的模糊综合评价等级,利用Isograph软件实现故障模式传递的可视化以及直观表达故障模式危害度矩阵,利用模糊综合评价的结论及易知部件的故障率求得边界阀的故障率。     

Passive measure for preventing recriticality of fuel debris dust for defueling at Fukushima Daiichi nuclear power station

The drilling or cutting of resolidified fuel debris required to defuel the Fukushima Daiichi nuclear power station is certain to generate debris dust. This paper focused on drilling resolidified fuel debris in water and conservatively confirmed by criticality calculations that neutron multiplication effect is higher if debris dust is suspended separately from the debris rather than if it is suspended closely around the debris. No use of vacuuming of debris dust, borated water, and active components was assumed in this study. Also, this paper confirmed that the use of a debris dust guide effectively and passively limited the increase in neutron multiplication by debris dust because the guide distributes dust particles so flatly that sufficient neutron leakage limits neutron multiplication even if the volume fraction of the particles in water reaches the optimum condition. In the actual defueling operation at the nuclear power station, the use of a flatter debris dust guide will be more effective to prevent local recriticality concurrently with the careful control of the mass of debris dust. The physics and ideas in this paper should be applicable to other defueling technologies such as laser cutting as long as debris dust is generated and suspended in water.     

基于SCALE的压水堆乏燃料贮存水池临界安全分析

基于国际先进的核设计与安全分析计算程序SCALE,针对我国自主研发的先进压水堆乏燃料贮存水池,建立恰当的计算模型,并选取合理的保守假设,计算乏燃料水池正常贮存及事故工况下的反应性,评估计算模型的临界安全,分析该程序对我国先进反应堆乏池计算的适用性。计算结果表明该先进压水堆乏燃料贮存水池正常贮存工况及事故工况的有效增值因子均小于0.95,处于次临界状态。该设计模型可确保燃料堆内贮存区域临界状态安全可控。SCALE计算程序适用于我国自主研发的先进压水堆乏燃料水池临界安全计算。     

秦山CANDU6重水堆应用RBGSS技术的可行性分析

RBGSS技术将为CANDU6重水堆提供另外一种进入保证停堆状态的方法。本文介绍了RBGSS的技术方案,分析了RBGSS技术的优势,结合秦山CNADU6反应堆的运行实践,探讨了RBGSS技术在机组大修和小修过程中的应用方法,评价了RBGSS技术的安全性和经济性。最后讨论了RBGSS可能存在的问题。     

塔里木油田氮气钻井内防喷器组的可靠性分析

综合应用故障模式影响及危害度分析（FMECA）与故障树分析(FTA)对氮气钻井内防喷器组进行可靠性分析。首先对氮气钻井内防喷器组进行系统级和设备级的故障模式影响及危害度分析,找出各个设备的常见故障模式,提出对应的改进措施,并定量计算出故障模式的危害度和系统内各个内防喷工具的危害度,为故障树分析确定顶事件。故障树分析,求出导致氮气钻井内防喷器组主要故障的所有底事件和最小割集,进而确定底事件的结构重要度的大小和底事件的组合方式对系统故障的影响力,对结构重要度大的底事件和底事件的组合方式对系统故障的影响力高的故障形式优先采取改进措施,并对薄弱零部件进行可靠度计算,更有针对性地对薄弱零部件进行设计改进和故障监测。鉴于对氮气钻井内防喷器组的可靠性研究,得出箭型止回阀锥面密封故障、翻板浮阀的复位弹簧失效、投入式止回阀的投球被压溃、旁通阀的扣断裂和旋塞阀的旋球被“抱死”是导致系统失效的主要故障模式,分析5种故障模式的所有可能的原因并提出了相应的改进方案。     

基于复杂网络理论的输电容量匹配模型

为提高电力系统供电可靠性,减少大规模连锁故障发生的几率,针对连锁故障的传播提出一种线路重要度评估指标,用于量化输电线路故障对于连锁故障传播的作用,并根据线路的重要度评估结果给出了输电容量匹配模型。藉由该匹配模型可以合理配置系统的输电容量,即优化系统的负载率分布从而达到提高系统供电可靠性的目的。仿真实验结果证明,在系统输电容量冗余,即系统整体负载率相同的情况下,所提出的优化模型匹配输电线路的输电容量或调整负载率分布可以防止故障传播过程中系统负载率向异质分布的演化趋势,从而降低大规模连锁故障发生的几率。     

核临界事故报警系统布置分析方法研究

本文阐述了布置核临界事故报警系统的意义和原则,分析了核临界事故可能发生的机理,初步建立了一套核临界事故情景假设分析方法。研究了最小临界事故源项计算方法以及三维剂量场分布计数的方法,采用各设备最小临界事故剂量场分布最小值等高线图的方法来从众多剂量场分布图中优化选取合适的核临界事故报警系统布置点位,以确保其可以覆盖到每个具有核临界事故风险的设备,并对核临界事故报警系统探头类型选择的原则和方法进行了分析。     

Monte Carlo criticality calculation under extreme condition

Autocorrelation and bias characteristics in the Monte Carlo criticality calculation have been examined for a certain type of extreme problems. Possibility is shown that the bias of the effective multiplication factor can become larger than ?0.1 in magnitude for population size as large as 10,000 starter particles per generation. This bias dramatically decreases if the number of starter particles per generation is large enough so that the noise propagation of a source distribution becomes linear. Furthermore, under this linearity, the bias correction is demonstrated to be an estimation problem of autocorrelation. Therefore, the linearity diagnosis of the noise propagation should be made available in the Monte Carlo criticality codes in public release in order to encompass analysis scenarios beyond normal circumstances.     

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.     

Analysis of Heavy-Water-Moderated,Cluster-Type Fuel Lattices by Cluster Physics Code ‘MESSIAH’

This paper presents a comprehensive analysis performed by a new cluster analysis code ‘MESSIAH’ on reactor physics constants measured in the critical facility for a pressure-tube-type, heavy-water-moderated reactor. The MESSIAH code system utilizes the method of the collision probability to solve the neutron transport equation. The effective space dependent cross sections are calculated in the thermal and resonance energy range before the eigenvalue calculation for the whole energy range. With use of these cross sections, the multi-group, space dependent transport equation is solved, and the flux distribution, spectrum and k _eff are obtained to the input bucking. In the above three steps the method of the collision probability is used consistently and extensively. The treatment of leakage neutrons from lattices in MESSIAH is also confirmed by an independent method using a Monte Carlo calculation. The calculated reactor physics constants, especially the micro-parameters and the activation traverse of Dy, agreed fairly well with the experiment. The diffusion calculation with use of the group constants calculated by MESSIAH predicts the reactivity of 0% void core excellently (<0.12%). However, for a 100% void core, the calculated reactivity was slightly lower than the experiment (~0.74%), which was attributed to over prediction of the diffusion constants.