Human reliability analysis in the man–machine interface design review

Advanced, computer-based man-machine interface (MMI) is emerging as part of the new design of nuclear power plants. The impact of advanced MMI on the operator performance, and as a result, on plant safety should be thoroughly evaluated before such technology is actually adopted in the plants. This paper discusses the applicability of human reliability analysis (HRA) to support the design review process. Both the first-generation and the second-generation HRA methods are considered focusing on a couple of promising HRA methods, i.e. ATHEANA and CREAM, with the potential to assist the design review process.     

An update on project COSA

The current status of a benchmark exercise to compare computer predictions of the thermal—creep behaviour of rock salt is described. Interest centres on the reliability of predictions of long term behaviour of repositories for high level (i.e. heat producing) waste in geological formations. Three benchmark problems are discussed and the results of one are presented in detail. An analysis of the differences in the predicted behaviour is attempted.     

适用于Living PSA的故障树求解方法

Living PSA是当前核电厂的安全分析与评价中最热点的问题之一，Living PSA实现中最为根本而又关键的问题是“速度”问题：在现有PSA方法的基础上，根据Living PSA的特性，设计了一种适用于Living PSA的故障树求解方法，即独立模块排序求解法，并通过例子详细地说明了该方法实现过程：这个方法除了能快速地求解故障树以外，当故障树结构或故障树中部件可靠性模型或数据发生变化时能实时地求解故障树。     

Statistical model for the random cyclic strain–life relations of 1Cr18Ni9Ti pipe-weld metal under temperature of 240°C

Modeling of random cyclic strain–life (CSL) relations of engineering material should be a basis of strain-based fatigue reliability analysis. A statistical model for the relations of a nuclear engineering material, 1Cr18Ni9Ti stainless steel pipe-weld metal under temperature of 240°C, is presented. In the model, a verified distribution, i.e. lognormal distribution, is used as an appropriate assumed distribution of the material fatigue life data. Based on the Coffin–Manson law, the relations are modeled by mean value- and standard deviation-cyclic curves of the logarithm of fatigue life. Then, fatigue analysis at an arbitrarily given probability can be made conveniently according to the normal distribution function. An approach for estimating the curves and their confidence bounds is developed by a linear regression technique. Different from the existent reliability analysis methods that considered the material constants in the law as independently random variables, present work treats them as dependently random variables from the fit of test data. Availability of the model has been indicated by an analysis of the material test data.     

SPAR-H方法中行为形成因子间的相关性识别

标准化核电厂风险分析-人因可靠性分析方法（SPAR-H）是目前国际上认可和接受的人因可靠性分析方法,但其8个行为形成因子（PSFs）间存在交叉部分,导致人因失误概率重复计算或高估。为了改进SPAR-H的PSFs体系,通过统计2007年到2017年219份国内核电厂运行事件报告,筛选出与主控室操纵员运行有关的89份人因事件/事故报告进行PSFs相关性的研究,运用数据挖掘技术（关联规则分析、探索性因子分析、皮尔森相关性分析）对统计结果进行分析。结果表明:①复杂度、压力、职责适宜以及可用时间4个PSFs之间存在相关性。其中,复杂度分别与压力和职责适宜相关,职责适宜与压力、压力与可用时间相关;②工作过程、规程、人因工程/人机界面和经验/培训之间存在关联。在涉及经验/培训、人因工程/人机界面和规程的事件中,很大概率还涉及到工作过程。这些结论可以给改进SPAR-H的PSFs体系提供参考,为定量研究PSFs间的因果关系建立基础。      

Reliability assessment of emergency exhaust system in a pool-type research reactor

The reliability of an extract system in a swimming-pool-type research reactor has been assessed. A global fault-tree analysis technique has been utilized. The basic event reliability data is based on both generic and reactor specific informations.The unavailability of the extract system is quantified in terms of the unavailability of the various functional requirements of the system. The unavailability is expressed as the probability of failure on demand. The computer system unavailability is determined from the minimal cutsets of the system. It is found that only three events have a major contribution to the top event, i.e., failures of compressed air supply, electric power supply and solenoid valve. A sensitivity analysis is performed to show the effects of variations in the data values of the dominant cutsets. An uncertainty analysis was also performed on the fault tree. The evaluations show that the reactor extract system lacks diversity and redundancy in most of its components. It is tolerant of most minor degradations, as these are taken care of by the operating policies and procedures. However, it can not tolerate common cause failures, e.g., simultaneous compressed air and electric power supply failure. Based upon the results obtained some recommendations are made.     

Reliability evaluation of a natural circulation system

This paper discusses a reliability study performed with reference to a passive thermohydraulic natural circulation (NC) system, named TTL-1. A methodology based on probabilistic techniques has been applied with the main purpose to optimize the system design. The obtained results have been adopted to estimate the thermal-hydraulic reliability (TH-R) of the same system.A total of 29 relevant parameters (including nominal values and plausible ranges of variations) affecting the design and the NC performance of the TTL-1 loop are identified and a probability of occurrence is assigned for each value based on expert judgment. Following procedures established for the uncertainty evaluation of thermal-hydraulic system codes results, 137 system configurations have been selected and each configuration has been analyzed via the Relap5 best-estimate code. The reference system configuration and the failure criteria derived from the “mission” of the passive system are adopted for the evaluation of the system TH-R.Four different definitions of a less-than-unity “reliability-values” (where unity represents the maximum achievable reliability) are proposed for the performance of the selected passive system. This is normally considered fully reliable, i.e. reliability-value equal one, in typical Probabilistic Safety Assessment (PSA) applications in nuclear reactor safety. The two ‘point’ TH-R values for the considered NC system were found equal to 0.70 and 0.85, i.e. values comparable with the reliability of a pump installed in an “equivalent” forced circulation (active) system having the same “mission.” The design optimization study was completed by a regression analysis addressing the output of the 137 calculations: heat losses, undetected leakage, loop length, riser diameter, and equivalent diameter of the test section have been found as the most important parameters bringing to the optimal system design and affecting the TH-R.As added values for this work, the comparison has been made between results from this study and results from a previous analysis where the same methodology was adopted for the evaluation of the TH-R of a different passive system named Isolation Condenser (IC). The comparison shows that the current single-phase NC system is ‘more reliable’ than the two-phase IC system. This constitutes a proof of qualification and of consistency for the adopted methodology.     

Control rod worth calculations using deterministic and stochastic methods

Knowledge of the efficiency of a control rod to absorb excess reactivity in a nuclear reactor, i.e. knowledge of its reactivity worth, is very important from many points of view. These include the analysis and the assessment of the shutdown margin of new core configurations (upgrade, conversion, refuelling, etc.) as well as several operational needs, such as calibration of the control rods, e.g. in case that reactivity insertion experiments are planned. The control rod worth can be assessed either experimentally or theoretically, mainly through the utilization of neutronic codes. In the present work two different theoretical approaches, i.e. a deterministic and a stochastic one are used for the estimation of the integral and the differential worth of two control rods utilized in the Greek Research Reactor (GRR-1). For the deterministic approach the neutronics code system SCALE (modules NITAWL/XSDRNPM) and CITATION is used, while the stochastic one is made using the Monte Carlo code TRIPOLI. Both approaches follow the procedure of reactivity insertion steps and their results are tested against measurements conducted in the reactor. The goal of this work is to examine the capability of a deterministic code system to reliably simulate the worth of a control rod, based also on comparisons with the detailed Monte Carlo simulation, while various options are tested with respect to the deterministic results’ reliability.     

Considerations of uncertainties in evaluating dynamic reliability by GO-FLOW methodology – example study of reliability monitor for PWR safety system in the risk-monitor system

The uncertainty analyses have been considered as a relevant topic since WASH-1400 and analysis was performed for identifying the risk measure, e.g. plant- and core-damage frequency or the frequency of a large early release of radioactivity in the probabilistic safety assessment (PSA) or probabilistic risk assessment. There are two main sources of uncertainty such as aleatory uncertainty and epistemic uncertainty (parameter uncertainty, model uncertainty and completeness uncertainty) for risk analysis in PSA or risk-monitor system. A sensitivity analysis is related field to uncertainty, which can provide information of the most effective on those inputs of PSA, which are mostly contributed to the uncertainty.

In this paper, uncertainty analysis (epistemic) has been conducted in the evaluation of dynamic reliability of safety-related subsystem for risk analysis. GO-FLOW methodology has been employed for the procedure of uncertainty analysis alternatively to Fault Tree Analysis and Even Tree because it is success-oriented system-analysis technique and comparatively easy to conduct the reliability analysis of the complex system. The method used sample data from Monte Carlo simulation to quantify uncertainty in terms of appropriate estimates for analysis results. Pressurized water reactor containment spray system has been taken as an example of safety-related subsystem. The results of this paper show that the uncertainty analysis is an important part for the practical evaluation of the system dynamic reliability and makes the reliability prediction more accurate compared with the result without the uncertainty analysis. The GO-FLOW methodology can be employed easily for uncertainty analysis with its advance functions.     

Feasibility study of the IE-SASW method for nondestructive evaluation of containment building structures in nuclear power plants

The IE-SASW method, a combination of impact-echo (IE) acoustics with spectral analysis of surface waves (SASW), is proposed as a newly developed nondestructive testing method in concrete structures. This feasibility study examines the IE technique and uses elastic P-wave velocity data as measured from the SASW method on concrete members in nuclear power plant containment structures. It was shown that both the thickness of the concrete specimens used in this study and the depth of the introduced defects (i.e. voids) could be identified by the IE-SASW method. In contrast, the reinforced steel bar itself could not be identified by the IE-SASW method. Additionally, GPR (ground penetrating radar) techniques were used to examine the same specimens in order to establish some level of performance and reliability to compare with the performance of the IE-SASW method. The GPR method provides an objective and reliable image corresponding to the reinforced steel bars. The experimental studies show that it is more feasible to use the IE-SASW method rather than GPR to detect voids that were positioned beneath the steel reinforcing bars in the concrete specimens.     

新产品开发中3F一体化技术的应用

3F一体化技术是由以FMECA与FTA综合分析方法为基础构成的潜在故障预测和改善功能模块及以FRACAS为原型的已知故障闭环控制功能模块组成的可靠性分析技术.描述了将3F一体化技术应用于新产品开发全过程的研究思路,重点阐述3F一体化技术在新产品开发过程中的运行模型及实施流程,并以案例演示其具体实施过程.最后,对3F一体化技术及其在新产品开发中的应用,提出了有待发展研究的课题.     

A complex study on the reliability assessment of the containment of a PWR Parts I–III

The intention of this paper is to contribute to the development of methods to be used for the quantification of the risk of nuclear power plants. For this purpose a reliability analysis of a structural component, i.e. a reactor containment structure is carried out. Detailed information in various fields had to be developed and compiled. The project consists of three parts: Part I concentrates mainly on the analysis of the load condition of the steel hull following a Loss of Coolant Accident (LOCA). Part II deals with the material aspects of the design properties of containment steels and furthermore the behaviour of concrete under impact load conditions are discussed. Part III of the paper is concerned on the one hand with external load conditions, and on the other hand with assembling the information of the previous parts to a reliability analysis.The methodology is exemplified by applying the general and theoretical results to the containment of the PWR-plant “Biblis B”.     

Reliability data update method for emergency diesel generator of Daya Bay Nuclear Power Plant

In the field of Living Probabilistic Safety Assessment (LPSA) the reliability data updating is an important factor. In risk analysis equipment failure data is needed to estimate the frequencies of events contributing to risk posed by a facility. Five years data of emergency diesel generator (EDG) of Daya Bay Nuclear Power Plant (NPP) has been studied in this paper. The data updating process has been done by using two methods, i.e., the classical method and Bayesian method. The aim of using these methods is to calculate the operational failure rate (λ) and demand failure probability (p). The results show that the operational failure rate is 1.7E?3 per hour and the demand failure probability is 2.4E?2 demand per day for Daya Bay NPP. By comparing the results obtain from classical and Bayesian methods with EDF (Electric De France) it is concluded that the design and construction of Daya Bay NPP is very different than EDF therefore the reliability parameters used in Daya Bay NPP is based on the classical method.     

基于响应面拟合方法中国铅基研究实验堆非能动余热排出系统可靠性分析

非能动系统已广泛地应用于新一代堆的设计中,其可靠性分析成为新型反应堆概率安全评价(Probabilistic Safety Analysis,PSA)的重要内容。本文提出一种用于非能动系统可靠性分析的响应面拟合方法,并应用于中国铅基研究实验堆反应堆容器空气冷却系统(Reactor Vessel Air Cooling System,RVACS)的可靠性分析。采用流体计算软件Fluent模拟RVACS系统的输入输出作为求解响应面性能函数的输入样本,利用最小二乘法和bootstrap方法估计响应面性能函数的系数,以响应面模型代替Fluent模型分析RVACS系统的非能动失效概率。分析表明,在所有能动余热排除系统不可用的情况下,RVACS四组并联排热管中的两组也能够可靠地导出反应堆余热。RVACS系统可靠性高。     

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.     

Extreme value risk analysis in the structural design of reactor components

The tendency to confuse “uncertainties” associated with design assumptions and parameters and compensated by the safety factor with objective ‘risks of failure’ implicit in the design, has been characteristic of the approach to probability-based structural design on all levels. However, a clear differentiation between uncertainty and risk is required to remove the lack of correlation between design safety analysis and risk analysis implicit in the present approach to the design of major structural and mechanical components of nuclear reactors as well as other structures.In a recent paper [5] the author has used the definition of the safety factor as a random variable (distribution of a quotient) to construct a probability model that justifies the introduction of the asymptotic distributions of extreme values as the physically relevant distributions of the design parameters governing ultimate load failure on which a realistic risk assessment can be based. Realistic reliability and risk assessment of reactor components subject to fatigue and creep, i.e. design conditions that exceed in practical importance that of ultimate load failure, can be based on the use of the third asymptotic distribution of smallest values.In the case of structural components working under complicated conditions it becomes necessary to perform full-scale tests reproducing, as closely as possible, the anticipated operational and, whenever necessary, critical limiting conditions to be provided in the design as well as in the associated reliability and risk assessment. The economic necessity of keeping the number of such full-scale tests to a minimum which, in the case of larger components, is usually a single or very small number of tests, raises the problem of integration of the test results into the framework of a reliability and risk assessment.     

Principles of reliability analysis methods applied to an emergency core cooling system

A reliability analysis is given for a model system of an emergency core cooling system of a pressurized water reactor. In order to demonstrate some basic relationships and influences on the system's failure probability the analysis deals only with some of the main-components and subsystems of the emergency core cooling system. With reference to the design basis accident, i.e. the total rupture of a main coolant line, only the low pressure system is considered. The overall system's failure probability is determined by the failure probability per demand, i.e. the unavailability of the system when called on for operation in the emergency case, and the cumulative probability of failure during the subsequent phase of residual heat removal. Detailed calculations have shown that the failure probability per demand is the leading term. Special attention is given to some parameter calculations dealing with the influence of inspection time intervals and repair procedures for different components and subsystems with respect to system failure behaviour.     

Integration of functional reliability analysis with hardware reliability: An application to safety grade decay heat removal system of Indian 500 MWe PFBR

A passive system can fail either due to classical mechanical failure of components, referred to as hardware failure, or due to the failure of physical phenomena to fulfill the intended function, referred to as functional failure. In this paper a methodology is discussed for the integration of these two kinds of unreliability and applied to evaluate the integrated failure probability of the passive decay heat removal system of Indian 500 MWe prototype fast breeder reactor (PFBR). The probability of occurrence of various system hardware configurations is evaluated using the fault tree method and functional failure probabilities on the corresponding configurations are determined based on the overall approach reported in the reliability methods for passive system (RMPS) project. The variation of functional reliability with time, which is coupled to the probability of occurrence of various hardware system configurations is studied and incorporated in the integrated reliability analysis. It is observed that this consideration of the dependence of functional reliability on time will give significant advantages on system reliability. The integrated reliability analysis is also explained using an event tree. The impact of the provision for forced circulation in the primary circuit on functional reliability is also studied with this procedure and it is found that the forced circulation capability helps to bring down the total decay heat removal failure probability by lowering the peak temperatures after the reactor shut down.     

The price of risk reduction: Optimization of test and maintenance integrating risk and cost

Testing and maintenance (T&M) improve the reliability of safety systems and components in nuclear power plants, which is of special importance for standby systems. Early optimizations of single component test intervals were based on minimizing the risk, e.g. the time-average unavailability, without cost considerations. However, the appropriate development of T&M strategy depends not only on the T&M intervals but also on the resources (human and material) available to implement such strategies. Since these testing and maintenance activities are associated with substantial cost, they present an important domain, where risk reduction and costs can be balanced.The objective of this paper focuses on assessing how costs and component ageing may affect the T&M optimization in terms of minimal system risk. The costs are expressed as a function of the selected risk measure. The time-averaged function of the selected risk measure is obtained from probabilistic safety assessment, i.e. the fault tree analysis at the system level, extended with inclusion of time parameters related to T&M activities. Additionally, component ageing is taken into account while developing the system reliability model presented in this paper. The testing strategy is also addressed. Sequential and staggered testing strategies are compared. The developed approach is applied on a standard test system and the obtained results are presented. The results show that the risk-informed surveillance requirements differ from existing ones in technical specifications, which are deterministically based. The presented approach achieves a significant reduction in system unavailability accompanied with relatively small changes in total T&M costs.     

Addressing the heavy liquid metal – Water interaction issue in LBE system

The interaction between heavy liquid metal (HLM) and water is a safety concern for the preliminary designs of lead fast reactor (i.e. LFR) and of subcritical transmutation system prototypes (i.e. XT-ADS). Current pool-type configurations have steam generators (SG) inside the reactor vessel. This implies that the primary to secondary leak (e.g. steam generator tube rupture) shall be considered as a postulated initiating event. The issue is addressed for CIRCE facility in ICE (Integral Circulation Experiment) configuration. CIRCE facility is a large pool system aimed at studying key operating principles of Lead Bismuth Eutectic (and Lead) systems. The configuration ICE was carried out to perform integral experiments, simulating the coupling between a high-performance heat source (electrically heated fuel bundle) and the heat exchanger, which was representative of the preliminary design of the XT-ADS heat exchanger. A Failure Mode and Effect Analysis (FMEA) is applied in order to get a complete picture of all the failure modes pertaining to this system, to determine their effects and to classify them according to their severity. The outcome of the analysis has identified as major hazard, relative to the CIRCE facility in the ICE configuration, the risk related to the LBE/water reaction, although with a very low probability, with the potential for a suddenly and dangerous pressurization (beyond the failure threshold) within the main vessel. A SIMMER-III code model of the system has been setup to provide deterministic results of the scenario. The results are supported by means of a LBE/water interaction experiment executed in LIFUS5 facility. LIFUS5 is a separate effect test facility dedicated to the investigation of LBE/water interaction. SIMMER-III code pre-test and post-test analyses are performed to define the boundary conditions of the experiment and to demonstrate the reliability of the code in simulating the phenomena of interest. The activity contributes to solving the safety issue raised for the operation of CIRCE facility and it provides a sample approach for addressing the safety studies needed in the development of the lead fast reactor and of the subcritical transmutation system.