Composition of Safety and Cyber Security Analysis Techniques and Tools for NPP I&C System Assessment

2020 ◽  
pp. 204-220
Author(s):  
Ievgen Babeshko ◽  
Kostiantyn Leontiiev
Keyword(s):  
Cyber Security ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Security Analysis ◽  
Assessment Method ◽  
Assessment Methods ◽  
Assessment Process ◽  
Safety Level ◽  
Methods And Techniques ◽  
And Control

Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource-consuming process that is required to be done so as to ensure the required safety level and comply to normative regulations. A lot of work has been performed in the field of application of different assessment methods and techniques, modifying them, and using their combinations so as to provide a unified approach in comprehensive safety assessment. Performed research has shown that there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. This chapter presents a developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs.

NPP I&C Safety Assessment by Aggregation of Formal Techniques

2018 ◽  
Author(s):  
Eugene Babeshko ◽  
Vyacheslav Kharchenko ◽  
Kostiantyn Leontiiev ◽  
Oleg Odarushchenko ◽  
Oleksiy Strjuk
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Failure Modes ◽  
Assessment Method ◽  
Assessment Methods ◽  
Assessment Process ◽  
Software Safety ◽  
Safety Level ◽  
Normative Documents

Safety assessment of nuclear power plant instrumentation and control systems (NPP I&Cs) is a complicated and resource consuming process that is required be done so as to ensure the required safety level and comply to normative regulations. A lot of work have been performed in the field of application of different assessment methods and techniques, modifying them and using their combinations so as to provide unified approach in comprehensive safety assessment. Anyway, performed research have shown there are still challenges to overcome, including rationale and choice of the safety assessment method, verification of assessment results, choosing and applying techniques that support safety assessment process, especially in the nuclear field. In our work we present developed framework that aggregates the most appropriate safety assessment methods typically used for NPP I&Cs. Key features that this framework provides are the formal descriptions of all required input information for every safety assessment method, possible data flows between methods, possible output information for every method. Such representation allows to obtain possible paths required to get necessary indicators, analyze the possibility to verify them by application of different methods that provide same indicators etc. During safety assessment of NPP I&Cs it is very important to address software due to its crucial role in I&C safety assurance. Relevant standards like IEC 60880 [1] and IEC 62138 [2] provide requirements for software related activities and supporting processes in the software safety lifecycle of computer-based I&C systems of nuclear power plants performing functions of safety category A, B and C, as defined by IEC 61226 [3]. Requirements and frameworks provided by IEC 60880 and IEC 62138 for the nuclear application sector correspond to IEC 61508, part 3 [4]. These standards define several types of safety related software and specify particular requirements for each software type. So as to verify software and confirm correspondence to required safety level, different techniques are suggested in normative documents. We share our experience obtained during software failure modes and effect analysis (software FMEA) and software fault insertion (software FIT) processes into FPGA-based platform, NPP I&C systems based on that platform, and RPCT, integrated development environment used by RPC Radiy and end users to design user application logic, specify hardware configuration etc. We apply software FIT to outputs of RPCT, considering source code, configuration files and firmware files. Finally, we provide a case study of application the developed safety assessment framework and software FMEA/FIT practices during practical assessment of FPGA-based NPP I&C system.

Cyber Security Assessment of NPP I&C Systems

2020 ◽  
pp. 221-238
Author(s):  
Oleksandr Klevtsov ◽  
Artem Symonov ◽  
Serhii Trubchaninov
Keyword(s):  
Control Systems ◽  
Cyber Security ◽  
Nuclear Power ◽  
Power Plants ◽  
Cyber Attacks ◽  
Security Assessment ◽  
Nuclear Facilities ◽  
Security Risks ◽  
Cyber Threats ◽  
And Control

The chapter is devoted to the issues of cyber security assessment of instrumentation and control systems (I&C systems) of nuclear power plants (NPP). The authors examined the main types of potential cyber threats at the stages of development and operation of NPP I&C systems. Examples of real incidents at various nuclear facilities caused by intentional cyber-attacks or unintentional computer errors during the maintenance of the software of NPP I&C systems are given. The approaches to vulnerabilities assessment of NPP I&C systems are described. The scope and content of the assessment and periodic reassessment of cyber security of NPP I&C systems are considered. An approach of assessment to cyber security risks is described.

scholarly journals Safety Assessment of Casting Workshop by Cloud Model and Cause and Effect–LOPA to Protect Employee Health

2020 ◽  
Vol 17 (7) ◽  
pp. 2555 ◽  
Author(s):  
Qingwei Xu ◽  
Kaili Xu ◽  
Fang Zhou
Keyword(s):  
Safety Assessment ◽  
Cloud Model ◽  
Least Square Method ◽  
Assessment Method ◽  
Employee Health ◽  
Least Square ◽  
Entropy Weight ◽  
Safety Level ◽  
Cause And Effect ◽  
Assessment Approach

Safety assessment of a casting workshop will provide a clearer understanding of the important safety level required for a foundry. The main purpose of this study was to construct a composite safety assessment method to protect employee health using the cloud model and cause and effect–Layer of Protection Analysis (LOPA). In this study, the weights of evaluation indicators were determined using the subjective analytic hierarchy process and objective entropy weight method respectively. Then, to obtain the preference coefficient of the integrated weight more precisely, a new algorithm was proposed based on the least square method. Next, the safety level of the casting workshop was presented based on the qualitative and quantitative analysis of the cloud model, which realized the uncertainty conversion between qualitative concepts and their corresponding quantitative values, as well as taking the fuzziness and randomness into account; the validity of cloud model evaluation was validated by grey relational analysis. In addition, cause and effect was used to proactively identify factors that may lead to accidents. LOPA was used to correlate corresponding safety measures to the identified risk factors. 6 causes and 19 sub-causes that may contribute to accidents were identified, and 18 potential remedies, or independent protection layers (IPLs), were described as ways to protect employee health in foundry operations. A mechanical manufacturing business in Hunan, China was considered as a case study to demonstrate the applicability and benefits of the proposed safety assessment approach.

Studies on Safety Assessment Method for Operating License Extension of Nuclear Power Plants in China

2017 ◽  
Author(s):  
Zhilin Chen ◽  
Ping Huang ◽  
Chunhui Wang ◽  
Zhiyuan Chi ◽  
Fangjie Shi ◽  
...  
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Assessment Method ◽  
Nuclear Safety ◽  
Management Systems ◽  
Effect Evaluation ◽  
Environment Effect ◽  
Assessment Environment

It’s the trend to extend the operating license time, called Operating License Extension (OLE) in China, of nuclear power plants (NPPs) in the future. It needs to be adequately demonstrated by licensees and approved by the regulator to gain an extended license time, such as 20 years. The demonstration methods for OLE are different among countries due to the different management systems for NPPs. Safety assessment, environment effect evaluation and update of the final safety analysis report (FSAR) will be the main aspects during OLE demonstration of NPPs in China according to the technical policy issued by National Nuclear Safety Administration (NNSA). Technical methods for scoping and screening, aging management review and time-limited aging analyses, which are the main contents of safety assessment are established based on the technical policy drafted by NNSA and international experiences in order to assist the operators to implement the safety assessment for OLE of NPP.

Research on Special Analysis of Verification and Validation in Nuclear Power Instrument and Control System

2017 ◽  
Author(s):  
Jingbin Liu ◽  
Yan Feng ◽  
Ning Qiao ◽  
Yunbo Zhang ◽  
Zhongqiu Wang
Keyword(s):  
Nuclear Power ◽  
Software Verification ◽  
Hazard Analysis ◽  
Security Analysis ◽  
Verification And Validation ◽  
Special Analysis ◽  
Life Cycle Stages ◽  
Traceability Analysis ◽  
And Control ◽  
Criticality Analysis

At present, there is still lack of detailed software V&V guidance standards in China, while a number of US nuclear power units and I&C platform are introduced and applied. So the software verification and validation work in our country usually cited the methods in IEEE 1012. With reference to the requirements of IEEE 1012, the V&V process of the software can be mainly divided into three forms: audit evaluation, special analysis and testing. This paper focuses on these parts and gives a detailed description and annotations of the technical methods and their life cycle stages in IEEE 1012, which cover multiple V&V phases. At the same time, the author puts forward his own understanding of the special analysis approach and procedure, such as criticality analysis, interface analysis, traceability analysis, hazard analysis, risk analysis and security analysis, and gives his own experience and related recommendations.

Application of Probabilistic Safety Assessment in the Design of Instrumentation and Control Systems for Nuclear Power Plants

2017 ◽  
Author(s):  
Pengyi Peng ◽  
Weidong Liu ◽  
Zhichao Yang
Keyword(s):  
Risk Assessment ◽  
Optimal Design ◽  
System Design ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Power Plants ◽  
Nuclear Power Plants ◽  
Probabilistic Safety Assessment ◽  
And Control ◽  
Probabilistic Safety

Instrumentation and control (I&C) systems in nuclear power plants (NPPs) have the ability to initiate the safety-related functions necessary to shut down the plants and maintain the plants in a safe shutdown condition. I&C systems of low reliability will bring risks to the safe operation of NPPs. A sufficient level of redundancy and diversity of I&C design to ensure the safety is a major focus when designing a new reactor. Usually multiple signal paths are included in an I&C system design. Meanwhile, besides the protection and safety monitoring system (PMS), other sub-systems of I&C such as the diverse actuation system (DAS) will be included as a diverse backup of PMS to perform the functions of reactor trip and engineered safety features actuation systems (ESFAS). However, the construction costs increase as the level of system redundancy and diversity grows. In fact, from the perspective of deterministic theory, an I&C system of only two chains can meet the single failure criterion. So how to obtain the balance of safety and economy is a challenging problem in I&C system designing. Probabilistic Safety Assessment (PSA) is the most commonly used quantitative risk assessment tool for decision-making in selecting the optimal design among alternative options. In this paper, PSA technique was used to identify whether the I&C system design offers adequate redundancy, diversity, and independence with sufficient defense-in-depth and safety margins in the design of a new reactor. Firstly, detailed risk assessment criteria for I&C design were studied and identified in accordance with nuclear regulations. Secondly, different designs were appropriately modeled, and the risk insights were provided, showing the balance of safety and economy of each design. Furthermore, potential design improvements were evaluated in terms of the current risk assessment criterion. In the end, the optimal design was determined, and uncertainty analyses were performed. The results showed that all four designs analyzed in this paper were met the safety goals in terms of PSA, but each design had a different impact on the balance of risk. As the support systems of the NPP we analyzed were relatively weak, loss of off-site power and loss of service water were two main risk contributors. The common cause failure of reactor trip breakers and the sensors of containment pressure were risk-significant. After identifying the major risk factors, the I&C design team can perform subsequent optimizations in the further design based on the PSA results and achieve an optimal balance between safety and economy.

The Application of Probabilistic Safety Assessment in the Preliminary Reactor Design Stage: Challenges and Insights

2014 ◽  
Author(s):  
Ibrahim A. Alrammah
Keyword(s):  
Nuclear Power ◽  
Safety Assessment ◽  
Design Stage ◽  
Human Reliability Analysis ◽  
Probabilistic Safety Assessment ◽  
Technical Issues ◽  
Technical Specifications ◽  
And Control ◽  
Novel Design ◽  
Probabilistic Safety

This paper discusses some technical issues related to applying Probabilistic Safety Assessment (PSA) to a novel Nuclear Power Plant (NPP) design. These aspects include: initiating events, passive systems modeling, reliability and common-cause failure (CCF) data, modeling of novel design features, modeling of preventive maintenance and technical specifications, human reliability analysis (HRA), modeling of instrumentation and control (I&C), external hazards, PSA supporting studies and interpretation of PSA results for new plants.

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