Visual Study of Flow Behavior in Upper Plenum during Simulated Reflood Phase of PWR-LOCA

Fuel temperature is an important parameter in reactor safety. However, temperatures of fuel sub-assemblies in power reactors usually cannot be directly measured. A means therefore has been devised to use heat removed by coolant from fuel sub-assemblies following a reactor trip to estimate the pre-trip average fuel temperatures and fuel-to-coolant heat transfer coefficients. Sub-assembly coolant mass flow and outlet temperature measurements are used to determine the heat removed. After accounting for the contribution from release of stored heat from non-fuel components and from decay heat, the contribution due to release of fuel-stored heat alone during the trip transient can be inferred. Average fuel temperatures and fuel-to-coolant heat transfer coefficients of the FBR “MONJU” at 45% thermal power were estimated using this energy balance approach. The trip-test data derived estimates compared favorably with plant simulation code calculated values.     

岭澳核电站3,4号机组反应堆保护系统概率安全评价结果校算

为了评估数字化仪表控制系统对核电厂安全的影响,以电厂停堆系统和专设安全设施驱动系统为例,参考西门子公司提供的故障树逻辑,对主泵流量低及功率量程中子通量高于整定值停堆故障和蒸汽发生器(SG)低-低水位和同一SG中主给水流量低故障进行了概率安全分析.分析中分别采用西门子公司提供的输入数据及通过失效率、试验时间以及β因子方法计算得到的数据,对西门子的分析结果进行了校算,在主要割集和失效概率上得到更为真实的结果.结果表明,考虑2种多样性的反应堆保护系统停堆I&C功能需求失效概率均值为5.5×10~(-8),符合分布式控制系统(DCS)合同中确定的可靠性目标值(1.0×10~(-7))和辅助给水电动泵驱动信号功能需求失效概率均值(5.21×10~(-6)与8.32×10~(-6)),也符合DCS合同中确定的可靠性目标值(1.0×10~(-5)).     

Design and modification of steam generator safety system of FBR MONJU

MONJU is a prototype fast breeder reactor (FBR) in Japan. The sodium–water reaction in the steam generator (SG) is one of the important safety assessment items for a sodium cooled reactor like MONJU. MONJU is equipped with hydrogen gas detectors for the small water leak detection, gas pressure gauges for the medium leak and sensors of rupture discs for the large leak. As a design basis accident, one tube failure then failure propagation of neighboring three tubes is assessed to verify the structural integrity of the secondary components. A latest evaluation method on the design margin against the overheating tube rupture showed that the present SG system had not an enough margin in the worst case. For improving the margin, it needs to shorten the time of the sodium–water reaction by earlier water leak detection in the SG and sooner water ejection from the SG tubes. Therefore, MONJU is now carrying out the following modification works: (1) addition of steam relief valves, (2) addition of a gas pressure gauge with changing the interlock logic and lowering the trigger level, (3) reducing the opening of the valves on the SG gas flow line to the dump tank because of earlier detection for the pressure rise. After this modification, the design margin of the SG system will be sufficiently improved.     

Creep rupture assessment for Level-2 PSA of a 2-loop PWR: accounting for phenomenological uncertainties

The Level-2 probabilistic safety assessment (PSA) of pressurized water reactors studies the possibility of creep rupture for major reactor coolant system components during the course of high pressure severe accident sequences.The present paper covers this technical issue and tries to quantify its associated phenomenological uncertainties for the development of Level-2 PSA.A framework is proposed for the formal quantification of uncertainties in the Level-2 PSA model of a PWR type nuclear power plant using an integrated deterministic and PSA approach.This is demonstrated for estimation of creep rupture failure probability in station blackout severe accident of a 2-loop PWR,which is the representative case for high pressure sequences.MELCOR 1.8.6 code is employed here as the deterministic tool for the assessment of physical phenomena in the course of accident.In addition,a MATLAB code is developed for quantification of the probabilistic part by treating the uncertainties through separation of aleatory and epistemic sources of uncertainty.The probability for steam generator tube creep rupture is estimated at 0.17.     

基于蒙特卡罗抽样的简化核电站SBO事故的风险指引安全裕度分析

风险指引的安全裕度是近十年来核电行业提出的新的安全理念。本文研究风险指引的安全裕度的计算框架和蒙特卡罗抽样方法下的风险指引的安全裕度定量化技术,并重点研究蒙特卡罗抽样方法下的核电站全厂断电（SBO）事故下的风险指引的安全裕度定量化技术。借鉴蒙特卡罗抽样次数估算方法和基于蒙特卡罗的可靠度计算方法,根据蒙特卡罗抽样方法下的风险指引的安全裕度的不确定度计算方法以及蒙特卡罗抽样次数的估算流程,计算得出在绝对误差小于001或相对误差小于5%时,两种不同误差方法选择时SBO事故的风险指引的安全裕度计算的抽样次数,并分别完成两个抽样次数下核燃料包壳失效概率均值和标准差定量化计算。计算结果表明,不同的抽样方法、不同的正态分布对核燃料包壳失效概率均值和标准差均有显著影响。     

Reliability model of the digital reactor protection system considering the repair time and common cause failure

Reliability of the digital reactor protection system (RPS) is intensively researched as it is designed and installed to ensure the safety and economy which can be measured respectively by the probability of failure on demand (PFD) and probability of spurious trip (PST). Meanwhile, by analyzing the failure modes of the digital RPS, the failure on demand and spurious trip are the two main modes that should be evaluated for the reliability of digital RPS. Therefore, this paper develops the PFD and PST calculation formulas considering the module repair time as the repair takes some time, and during the repair duration, the digital system is operated in the degraded configuration and the common cause failure (CCF) which would severely impact the system in the event of occurrence. Considering the failure phenomenon of the digital RPS, the binomial failure rate (BFR) model is adopted for CCF. And the fault-tolerance techniques and their fault coverage are considered when calculating the PFD and PST. The quantitative results show that, in the example, CCF dominates the PFD while CCF is one of the major factors that result in PST but the main contributor is the independent failure. Also it can be concluded that the discovery time for the undetected failures dominates the PFD and PST when it costs long time to discover the failures even though the uncovered failures are very few. Thus, the failures should be covered by the fault-tolerance techniques as much as possible when designing the digital RPS.     

多模块高温气冷堆地震概率安全分析关键技术研究

高温气冷堆核电厂采取多个反应堆模块匹配1个汽轮机的设计方式,即1台高温气冷堆机组会包含多个反应堆模块,这使多个高温气冷堆模块在地震外部事件下存在明显的相关性,因此在利用概率风险分析方法来全面地识别和评价高温气冷堆的地震风险时,需要从机组的角度充分考虑和模化机组内多个反应堆模块间的相关性。高温气冷堆示范电站已完成了较为完整的单模块地震概率安全分析,本文将以该分析结果为基础梳理出高温气冷堆多模块地震概率安全分析的关键技术要素并进行研究,研究内容包括多模块事件序列建模和地震相关性失效评价等关键技术,并针对多模块高温气冷堆提出了应用策略。然后以双模块设计的高温气冷堆示范电站为对象,以地震导致丧失厂外电始发事件为代表,对多模块高温气冷堆地震概率安全分析进行了实例分析获得远低于概率安全目标的释放类频率,且分析得到了高温气冷堆多模块事件序列建模策略与地震相关性失效的评价路线可行这一重要结论。     

A model for estimation of reactor spurious shutdown rate considering maintenance human errors in reactor protection system of nuclear power plants

Modeling of spurious activations in safety instrumented systems has been studied for over a decade. The spurious activation of a plant protection system in nuclear power plants (NPPs) leads to increased electricity generation cost. An in-depth view on spurious activation of digital plant protection systems of NPPs for human errors in maintenance tasks is presented in this paper. A new model which considers human errors in maintenance and periodic tests to predict component failure rates is presented. The model has been applied to OPR-1000 reactor protection system for quantification of spurious trip frequency by fault-tree analysis. The major causes of spurious activation in a nuclear reactor protection system are identified. A set of case studies has been performed with the variation of magnitudes of human errors probability and maintenance strategies, in which, the human errors in maintenance are found to significantly influence reactor spurious trip frequency. This study is expected to provide a useful mean to designers as well as maintainers of the digital reactor protection system to improve plant availability and safety.     

田湾核电站安全仪控系统(TXS系统)失效概率估算

介绍了田湾核电厂安全仪控系统(TXS系统)失效概率的估算方法,推导了用于反应堆停堆系统和ESFAS系统失效概率估算的一般性公式。并以主给水/主蒸汽系统故障停堆和触发应急给水系统启动两个仪控功能为例进行计算,结果证明田核电站TXS系统失效概率满足可靠性要求。     

液态金属反应堆湿式燃料贮存桶内部事件概率安全分析

液态金属冷却剂在给反应堆带来运行安全与热效率优势的同时,也给反应堆带来了复杂的换料系统,其中大型液态金属反应堆采用的湿式乏燃料贮存桶是乏燃料卸料过程的核心设备,临时装载了大量的乏燃料组件,具备一定的安全风险。本文采用概率安全分析（PSA）方法对乏燃料贮存桶进行风险评价,通过运行状态分析、始发事件分析、事故序列分析以及简单的定量化,初步获得其导致乏燃料组件发生损伤的事故序列和最小割集,识别了关键系统与设备。结果表明,相对于反应堆本身的风险,乏燃料贮存桶本身风险虽低但依然不可忽略,且风险评价结果对反应堆的运行方式以及清洗系统的可靠性较为敏感。此外还对该系统的设计改进与安全优化进行了讨论。     

The use of the uniform risk spectra in the seismic PSA for a PWR power station

This paper describes a simple method for incorporating the effects of the uniform risk spectra (URS) in the seismic probabilistic safety assessment (PSA) for a pressurized water reactor (PWR) power station. The “traditional” fragility parameters for a range of critical equipment items in a PWR power station on two typical UK sites are modified to incorporate the URS using this simple method and the effect on the high confidence low probability of failure (HCLPF) acceleration levels and seismic-induced failure probabilities of the equipment items is examined. The results illustrate the potential benefit of using the URS in the seismic PSA for a PWR power station.     

冷却剂流量降低停堆保护系统整定值分析

在确保反应堆安全的基础上 ,尽量扩大电厂的运行区域是反应堆停堆保护系统设计以及整定值确定的原则。本文通过对电网运行要求的分析 ,得到了恰希玛核电厂主泵低转速和一回路低流量停堆整定值 ,随后的安全验证表明了其对冷却剂流量降低事故保护的有效性     

贝叶斯估计方法在Alpha因子模型参数估计中的应用

核电厂的概率安全分析(PSA)结果表明,共因失效(CCF)在系统的不可靠度中占有相当重要的贡献。国内PSA分析中CCF数据一直采用通用数据,难以体现国内核电机组的运行特点。Alpha因子模型由于其参数估计的简单化、计算结果的精确性等特点是PSA中最常用于模化共因失效的模型。但由于共因失效事件的罕见性,使用经典估计算法难以产生合理的统计值,因此,本文给出共因参数的贝叶斯估计方法,该方法能够结合先验信息和样本信息,不需要很大的样本就能得到较好的估计值,有效解决了核电厂共因失效事件少、使用经典估计方法计算结果不合理的问题,适用于核电厂共因失效模型参数估计。     

Numerical computation of fragility curves for NPP equipment

The seismic probabilistic risk assessment (PRA) methodology is a popular approach for evaluating the risk of failure of engineering structures due to earthquake. In this framework, fragility curves express the conditional probability of failure of a structure or component for a given seismic input motion parameter A, such as peak ground acceleration (PGA) or spectral acceleration. The failure probability due to a seismic event is obtained by convolution of fragility curves with seismic hazard curves. In general, a log-normal model is used in order to estimate fragilities. In nuclear engineering practice, these fragilities are determined using safety factors with respect to design earthquake. This approach allows to determine fragility curves based on design study but largely draws on expert judgement and simplifying assumptions. When a more realistic assessment of seismic fragility is needed, simulation-based statistical estimation of fragility curves is more appropriate. In this paper, we will discuss statistical estimation of parameters of fragility curves and present results obtained for a reactor coolant system of nuclear power plant. We have performed non-linear dynamic response analyses using artificially generated strong motion time histories. Uncertainties due to seismic loads as well as model uncertainties are taken into account and propagated using Monte Carlo simulation.     

核电站反应堆保护系统故障对策分析与应用

核电站反应堆保护系统应设计为当其任何部分出现故障均能保证反应堆的安全。根据反应堆停堆系统及专设安全设施驱动系统对故障安全的设计要求,研究了应对单一故障及共因故障的对策,并根据二代反应堆堆型的特点,设计了保护系统的基本架构。该架构的停堆系统采用2oo4表决逻辑,专设安全设施驱动系统采用2oo3表决逻辑,并提出了在输入发生失效时,表决逻辑的降级规则。     

Multiple aspects of raw data collection and processing for Romanian TRIGA SSR

Probabilistic Safety Assessment (PSA) is a tool for evaluating and enhancing the safety of a nuclear reactor. In general, PSA is used to support the system design, configuration decisions and the operational safety management of the plant. Ideally, failure data used for safety and reliability analyses should be based on site-specific data. The paper presents activities and results that started in 2001 with the participation of Romanian TRIGA Steady State 14 MW reactor in an international project for expanding component reliability database for Research Reactors for PSA use. Collection of reliability data for Romanian TRIGA is discussed emphasizing the problems encountered. Failure data for some components is given, both stand-alone and in comparison with other research reactors, as well as statistics pertaining to the failures and failure modes of the investigated components. A brief presentation of a software application developed in the Institute for Nuclear Research (INR) for raw data acquisition and processing is included.     

Concept for comprehensive Hazards PSA and Fire PSA application

This contribution presents results of recent research and development activities in the field of Hazards PSA (HPSA). The reactor accidents at Fukushima Dai-ichi in March 2011 gave reason and indications for checking the risk assessment approach for internal and external hazards as currently described in the German PSA Guideline and its supplementary technical documents. A standardized approach for performing a comprehensive HPSA has been developed emphasizing the complete consideration of all potential failure dependencies induced by hazards. The systematic extension of the given plant model of Level 1 PSA is the real crux of the new HPSA approach. The extension is carried out for each hazard H using the corresponding hazard equipment list (H-EL) and the corresponding hazard dependency list (H-DL). Parts of the approach have already been tested.In the paper a successful application for the plant internal hazard fire is presented. A German licensee plans a system modification of the spent fuel pool cooling, therefore a Level 1 PSA has been carried out to compare the fuel damage frequencies for the existing and the modified version. It is outlined how the systematic (and partly automatic) extension of the fault trees is performed using a so-called Fire Equipment List (F-EL). The F-EL contains a compartment assignment for all relevant components and cables. The probability of a compartment failure by fire must be determined for any compartment mapped. This is the conditional probability that the components and cables within the compartment are inoperable due to the fire.     

Incorporation of passive components aging into PRAs

The probabilistic risk assessments being developed at most nuclear power plants to calculate the risk of core damage generally focus on the possible failure of active components. The possible failure of passive components is given little consideration. We are developing a method for selecting risk-significant passive components and including them in probabilistic risk assessments. We demonstrated the method by selecting a weld in the auxiliary feedwater system. The selection of this component was based on expert judgement of the likelihood of failure and on an estimate of the consequence of component failure to plant safety. We then used the PRAISE computer code to perform a probabilistic structural analysis to calculate the probability that crack growth due to aging would cause the weld to fail. The calculation included the effects of mechanical loads and thermal transients considered in the design and the effects of thermal cycling caused by a leaking check valve. We modified an existing probabilistic risk assessment (NUREG-1150 plant) to include the possible failure of the auxiliary feedwater weld, and then we used the weld failure probability as input to the modified probabilistic risk assessment to calculate the change in plant risk with time. The results showed that if the failure probability of the selected weld is high, the effect on plant risk is significant. However, this particular calculation showed a very low weld failure probability and no change in plant risk for the 48 years of service analyzed. The success of this demonstration shows that this method could be applied to nuclear power plants.     

Passive system reliability in the nuclear power plants (NPPs) using statistical modeling

The probabilistic safety assessment (PSA) has been studied for the very high temperature reactor (VHTR). There is a difficulty to make the quantification of the PSA due to the deficiency of the operation and experience data. So, it is necessary to use the statistical data for the basic event. The physical data of the non-linear fuzzy set algorithm are used to quantify the designed case. The mass flow rate in natural circulation is investigated. In addition, the potential energy in the gravity, the temperature and pressure in the heat conduction, and the heat transfer rate in the internal stored energy are investigated. The values in the probability set and fuzzy set are compared for the failure explanation. The result shows how to use the newly made probability of the failure in the propagations. The failure frequencies, which are made by the GAMMA (GAs Multi-component Mixture Analysis) code, are compared with four failure frequencies by probabilistic and fuzzy methods. The results show that the artificial intelligence analysis of the fuzzy set could improve the reliability method than that of the probabilistic analysis.     

Fuzzy probability on reliability study of nuclear power plant probabilistic safety assessment: A review

Fault tree analysis (FTA) is a graphical model which has been widely used as a deductive tool for nuclear power plant (NPP) probabilistic safety assessment (PSA). The conventional one assumes that basic events of fault trees always have precise failure probabilities or failure rates. However, in real-world applications, this assumption is still arguable. For example, there is a case where an extremely hazardous accident has never happened or occurs infrequently. Therefore, reasonable historical failure data are unavailable or insufficient to be used for statistically estimating the reliability characteristics of their components. To deal with this problem, fuzzy probability approaches have been proposed and implemented. However, those existing approaches still have limitations, such as lack of fuzzy gate representations and incapability to generate probabilities greater than 1.0E-3. Therefore, a review on the current implementations of fuzzy probabilities in the NPP PSA is necessary. This study has categorized two types of fuzzy probability approaches, i.e. fuzzy based FTA and fuzzy hybrid FTA. This study also confirms that the fuzzy based FTA should be used when the uncertainties are the main focus of the FTA. Meanwhile, the fuzzy hybrid FTA should be used when the reliability of basic events of fault trees can only be expressed by qualitative linguistic terms rather than numerical values.