Safety management in NPPs using an evolutionary algorithm technique

The general goal of safety management in Nuclear Power Plants (NPPs) is to make requirements and activities more risk effective and less costly. The technical specification and maintenance (TS&M) activities in a plant are associated with controlling risk or with satisfying requirements, and are candidates to be evaluated for their resource effectiveness in risk-informed applications. Accordingly, the risk-based analysis of technical specification (RBTS) is being considered in evaluating current TS. The multi-objective optimization of the TS&M requirements of a NPP based on risk and cost, gives the pareto-optimal solutions, from which the utility can pick its decision variables suiting its interest. In this paper, a multi-objective evolutionary algorithm technique has been used to make a trade-off between risk and cost both at the system level and at the plant level for loss of coolant accident (LOCA) and main steam line break (MSLB) as initiating events.     

Regulation of nuclear power plants a multi objective approach

The regulation of nuclear power plant (NPP) is evolving in a direction to harmonize probabilistic safety criteria in the near future. The utilities will not only have to demonstrate that they are operating below a target risk level but also to demonstrate that the unavailability of some of the critical safety systems are below a specified level. In order to satisfy the Technical Specification and Maintenance (TS&M) requirements in a cost effective manner multi-objective optimization of TS&M requirements is of profound interest. The constrained multi-objective optimization of the TS&M requirements of a nuclear power plant (NPP) based on risk and cost gives the pareto-optimal solutions, from which the utility can pick suitable decision variables. The paper presents a multi objective genetic algorithm (GA) technique to investigate a trade-off between risk and cost both at the system and the plant level for Loss of Coolant Accident (LOCA) and Main Steam Line Break (MSLB) as initiating events in a NPP.     

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.     

基于风险指引安全分级的维修规则实施方案

近年来,美国核电厂的业绩始终保持世界领先水平,维修规则的实施起了很大的作用.本文研究了美国核电厂实施维修规则的法规要求以及实施方法,结合我国正在研究中的风险指引安全分级及其处理方法,提出了适用于我国的核电厂维修规则实施方案.     

Development of a standard communication protocol for an emergency situation management in nuclear power plants

Correct communication between main control room (MCR) operators is an important factor in the management of emergency situations in nuclear power plants (NPPs). For this reason, a standard communication protocol for the management of emergency situations in NPPs has been developed, with the basic direction of enhancing the safety of NPPs and the standardization of communication protocols. To validate the newly developed standard communication protocol, validation experiments with 10 licensed NPP MCR operator teams was performed. From the validation experiments, it was found that the use of the standard communication protocol required more time, but it can contribute to the enhancement of the safety of NPPs by an operators’ better grasp of the safety-related parameters and a more efficient and clearer communication between NPP operators, while imposing little additional workloads on the NPP MCR operators. The standard communication protocol is expected to be used to train existing NPP MCR operators without much aversion, as well as new operators.     

Development of CFD fire models for deterministic analyses of the cable issues in the nuclear power plant

Additional fire barriers of electrical cables are required for the nuclear power plants (NPPs) in Taiwan due to the separation requirements of Appendix R to 10 CFR Part 50. The risk-informed fire analysis (RIFA) may provide a viable method to resolve these fire barrier issues. However, it is necessary to perform the fire scenario analyses so that RIFA can quantitatively determine the risk related to the fire barrier wrap. The CFD fire models are then proposed in this paper to help the RIFA in resolving these issues. Three typical fire scenarios are selected to assess the present CFD models. Compared with the experimental data and other model’s simulations, the present calculated results show reasonable agreements, rendering that present CFD fire models can provide the quantitative information for RIFA analyses to release the cable wrap requirements for NPPs.     

风险指引的CPR1000核电厂LBLOCA分析方法初步研究

随着核电厂安全分析方法的不断发展,结合传统确定论分析与概率风险评价（PSA）的风险指引型安全分析方法逐渐引起安审当局和核电业主的广泛关注。本文基于国际上风险指引型分析方法在其他领域的应用现状,提出了风险指引的大破口失水事故（LBLOCA）分析方法,并重新评估了CPR1000核电厂的堆芯燃料包壳峰值温度（PCT）裕量。在PSA分析中,识别并量化了LBLOCA发生后可能发生的162个事件序列,并采用确定论现实分析方法（DRM）对筛选出的18个概率较大的事件序列进行了计算分析。然后通过期望值评估法和特定序列覆盖法对LBLOCA的PCT裕量进行了评估。结果表明,本文方法下LBLOCA的PCT裕量约为36~55 ℃,相比于传统的DRM裕量提升了16~35 ℃。     

Risk-informed inservice inspection program

The Nuclear Regulatory Commission (NRC) has developed draft guidance for power reactor licenses on acceptable methods for using probabilistic risk assessment (PRA) information and insights in support of plant-specific applications to change the current licensing basis (CLB) for inservice inspection (ISI) of piping. This process is also known as risk-informed inservice inspection programs (RI-ISI). The risk-informed inservice inspection process for operating nuclear power plants provides an alternative method for selecting and categorizing piping components that are inspected for the purposes of meeting the requirements of ASME Section XI. A RI-ISI approach will incorporate probabilistic techniques to help define the scope, type and frequency of inservice inspection. The risk-informed process may support a decrease in the number of inspection and inspection intervals but will also identify areas where increased resources should be allocated to enhance safety. The approach discussed in this paper follows the method developed by NRC staff.     

Risk managed technical specifications

Application of probabilistic risk assessment (PRA) technology has become an essential component in the decision-making processes associated with the operation and regulation of commercial nuclear power plants (NPPs). As PRA technology has matured, it increasingly has been utilized to provide risk insights in the support of both operational and regulatory decision-making. This paper describes the next significant application of PRA technology to risk inform NPP operation. This Risk Managed Technical Specification (RMTS) application utilizes the results of the plant PRA to determine risk-informed technical specification (TS) allowed out of service times (AOTs). The RMTS process utilizes the PRA results to specify appropriate configuration specific TS AOTs and ensures the risk of events that could result in core damage or large early release are maintained below acceptable levels. In addition, RMTS requires development of integrated risk management actions to actively mitigate risks associated with the inoperability of TS structures, systems and components (SSCs). RMTS has been approved for implementation at commercial NPPs in the United States with the South Texas Project Electric Generating Station (STPEGS) serving as the initial application. In this paper we describe the programmatic requirements necessary to implement RMTS and provide several examples illustrating its application; thus demonstrating the applicability of RMTS to manage nuclear safety risk while simultaneously enhancing operational flexibility.     

Reliability assessment method for NPP digital I&C systems considering the effect of automatic periodic tests

Since digital technologies have been improved, the analog systems in nuclear power plants (NPPs) have been replaced with digital systems. Recently, new NPPs have adapted various kinds of digital instrumentation and control (I&C) systems. Even though digital I&C systems have various fault-tolerant techniques for enhancing the system availability and safety compared to conventional analog I&C systems, the effects of these fault-tolerant techniques on system safety have not been properly considered yet in most probabilistic safety assessment models. Therefore, it is necessary to develop the safety evaluation method for digital I&C systems with consideration of fault-tolerant techniques. Among the various issues in the safety model for digital I&C systems, one of the important issues is how to exclude the duplicated effect of fault-tolerant techniques implemented at each hierarchy level of the system. The exact relation between faults and fault-tolerant techniques should be identified in order to exclude this duplicated effect. In this work, the relation between faults and fault-tolerant techniques are identified using fault injection experiments. As an application, the proposed method was applied to a module of a digital reactor protection system.     

Degradation assessment of structures and passive components at nuclear power plants

This paper describes a multi-year research program to assess age-related degradation of structures and passive components important to the safe operation of nuclear power plants (NPPs). The purpose of the research effort is to develop the technical basis for the validation and improvement of analytical methods and acceptance criteria which can be used to make risk-informed decisions and to address technical issues related to degradation of structures and passive components. The approach adopted for this research program consists of two phases. In Phase I, specific degradation occurrences at plants were collected and evaluated, existing technical information on aging was reviewed, and a scoping study was performed to identify which structures and components should be studied in the subsequent phases of the research program. Based on the results of the Phase I effort, selected structures and passive components are evaluated in Phase II to assess the effects of age-related degradation using existing and enhanced analytical methods. Fragility analyses are performed for undegraded and degraded structures and passive components. These results can then be used to assess the potential impact of degradation on overall plant risk. The Phase II effort also utilizes the results of the analyses to develop probabilistic degradation acceptance criteria for the structures and passive components studied. These research activities provide useful tools to support the current goals of developing risk-informed and performance-based regulation in the nuclear industry.     

Single event upset mitigation techniques for FPGAs utilized in nuclear power plant digital instrumentation and control

Field programmable gate arrays (FPGAs) are integrated circuits being increasingly used for digital instrumentation and control (I&C) in nuclear power plants (NPPs) because of low cost, re-configurability and low design turn-around time. However, to ensure reliability, proper design techniques must be employed since the memory and logic in FPGAs are susceptible to single event upsets (SEUs). Triple modular redundancy (TMR) has become a common SEU mitigation design technique because of its straightforward implementation and reliable results. Partitioned TMR approaches are introduced in this paper, and formulae derived indicate that the maximum probability of two simultaneous errors [P_E]_max is inversely proportional to the number of logic partitions in a TMR design, when each redundant logic block in every logic partition has the same number of sensitive nodes. However, the maximum logic partitioning design cannot completely eliminate the possibility of two simultaneous upsets. For the example test circuit it is found that [P_E]_max is reduced dramatically from 66.67% for minimum logic partitioning to 4.44% for maximum logic partitioning. Because TMR introduces significant overhead due to its full hardware redundancy, a dual modular redundancy approach is also examined for application to less demanding situations. By comparative analysis this study reaches the conclusion that the maximum logic partitioning TMR implementation is the best solution for digital I&C applications in NPPs where obtaining robustness is of the highest importance, despite its higher area overhead.     

风险指引型在役检查优化申请的独立审核计算

本文详细介绍了环境保护部核与辐射安全中心针对风险指引型在役检查(RI-ISI)优化申请所开展的独立审核计算。在大亚湾核电厂RI-ISI优化申请的审评工作中,采用国家核安全局标准概率安全分析(PSA)监管模型,计算管段失效后果和拟实施变更后的风险增量,对申请者管段失效后果分析结果进行核算,并独立评价该申请是否满足风险可接受准则。实现了核安全监管部门对PSA应用试点项目的独立审核计算,为核安全决策提供进一步的支持,提高核安全监管的独立性、科学性和有效性。     

Important factors affecting fault detection coverage in probabilistic safety assessment of digital instrumentation and control systems

As digital instrumentation and control (I&C) systems are gradually introduced into nuclear power plants (NPPs), concerns about the I&C systems’ reliability and safety are growing. Fault detection coverage is one of the most critical factors in the probabilistic safety assessment (PSA) of digital I&C systems. To correctly estimate the fault detection coverage, it is first necessary to identify important factors affecting it. From experimental results found in the literature and the authors’ experience in fault injection experiments on digital systems, four system-related factors and four fault-related factors are identified as important factors affecting the fault detection coverage. A fault injection experiment is performed to demonstrate the dependency of fault detection coverage on some of the identified important factors. The implications of the experimental results on the estimation of fault detection coverage for the PSA of digital I&C systems are also explained. The set of four system-related factors and four fault-related factors is expected to provide a framework for systematically comparing and analyzing various fault injection experiments and the resultant estimations on fault detection coverage of digital I&C systems in NPPs.     

Reliability-based service life assessment of concrete structures in nuclear power plants: optimum inspection and repair

Aging effects in reinforced concrete structures brought on by severe service conditions may cause their structural capacities to deteriorate gradually over their service life. Research is being conducted to address aging management of safety-related reinforced concrete structures in nuclear power plants (NPPs). Documentation is being prepared to identify potential structural safety issues and to recommend criteria for use in evaluating reinforced concrete structures for continued service. Time-dependent reliability analysis provides the framework and quantitative tools for the condition assessment. The role of in-service inspection and repair in ensuring continued reliability in service is examined. Optimum strategies can be determined on the basis of minimum life-cycle cost.     

System-level hazard analysis using the sequence-tree method

A system-level PHA using the sequence-tree method is presented to perform safety-related digital I&C system SSA. The conventional PHA involves brainstorming among experts on various portions of the system to identify hazards through discussions. However, since the conventional PHA is not a systematic technique, the analysis results depend strongly on the experts’ subjective opinions. The quality of analysis cannot be appropriately controlled. Therefore, this study presents a system-level sequence tree based PHA, which can clarify the relationship among the major digital I&C systems. This sequence-tree-based technique has two major phases. The first phase adopts a table to analyze each event in SAR Chapter 15 for a specific safety-related I&C system, such as RPS. The second phase adopts a sequence tree to recognize the I&C systems involved in the event, the working of the safety-related systems and how the backup systems can be activated to mitigate the consequence if the primary safety systems fail. The defense-in-depth echelons, namely the Control echelon, Reactor trip echelon, ESFAS echelon and Monitoring and indicator echelon, are arranged to build the sequence-tree structure. All the related I&C systems, including the digital systems and the analog back-up systems, are allocated in their specific echelons. This system-centric sequence-tree analysis not only systematically identifies preliminary hazards, but also vulnerabilities in a nuclear power plant. Hence, an effective simplified D3 evaluation can also be conducted.     

核电厂风险指引管理研究

风险指引管理是确定论与概率论方法相结合的一种新的安全管理模式.为了促进我国这项工作的开展,有必要对国内外的相关法规、标准和实践进行全面和系统的研究.本文介绍了核电厂风险指引决策的基本原则、方法与风险接受准则,讨论了风险指引决策对概率安全评价(PSA)的要求,并对我国核电厂采用风险指引管理提出了建议.     

风险指引的核电厂应急柴油发电机返厂大修分析

为了支持田湾核电站1、2号机组的1台应急柴油发电机进行返厂大修,本文提出了1台应急柴油发电机不可用的恢复时间的优化方案,并采用确定论与概率论相结合的风险指引型方法对优化方案的可行性进行了分析。分析结果表明,1台应急柴油发电机不可用的恢复时间优化满足相关法规导则和传统工程分析的要求,且对电厂带来的风险是很小的、可接受的。因此,可以开展1台应急柴油发电机进行返厂大修工作。      

GFR equilibrium cycle analysis with the EQL3D procedure

Advanced fast reactors of the fourth generation should be capable to breed their own fuel from ²³⁸U feed and to recycle the actinides from their own spent fuel. This recycling or virtually the closure of fuel cycle can converge to an equilibrium fuel cycle and has impact on the safety-related parameters. The goals of this study are: (i) to apply an equilibrium cycle procedure EQL3D to the Gas cooled Fast Reactor (GFR), (ii) to simulate and confirm the GFR neutronics capability for closed fuel cycle, and (iii) to evaluate the safety-related parameters of the equilibrium cycle.Equilibrium cycle method for considering the homogeneous recycling of actinides is a known approach. However, in EQL3D the equilibrium method is newly applied for hexagonal-z 3D core geometry and 33 energy-groups neutron-flux calculation. This geometry enables to characterize the equilibrium cycle for complex reloading patterns within a multi-batch cycle.Two GFR geometries were studied, the first based on an international neutronics benchmark with a simple set-up and the second based on more advanced core design. For the advanced design, three reloading patterns within a multi-batch cycle with four different feeds were compared.The GFR neutronics capability for closed cycle was proved. The negative impact of the fuel cycle closure on safety-related parameters was confirmed and quantified. The GFR core with closed fuel cycle could serve after prospective optimization as a sustainable and clean energy source.