Criticality analysis for safety-critical software in nuclear power plant distributed control system

Kerntechnik ◽  
2021 ◽  
Vol 86 (5) ◽  
pp. 343-352
Author(s):  
J. Cui ◽  
Y. Cai ◽  
Y. Wu
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Application Process ◽  
Software Module ◽  
Critical System ◽  
Safety Critical ◽  
Plant Safety ◽  
Safety Integrity Level ◽  
Criticality Analysis

Abstract Software criticality analysis examines the degree of contribution that each individual failure mode of a software component has on the reliability of software. Higher safety integrity levels are assigned to software modules whose failures cause an unacceptable impact on the operation of the system, and these levels require the implementation of more rigorous software quality assurance measures as defined in IEEE Std 1012 and in the customer’s system requirements specification. In this paper, a novel software criticality analysis method is proposed, the results of which can be used to guide the development of newly developed software and the procurement of Commercial-Off-The-Shelf (COTS) software. The software structure is first analyzed and the software is divided into modules according to their functions. Then the criticality levels of software components are preliminarily classified by means of a safety criticality preliminary analysis tree, followed by their verification through the software hazard and operability analysis (HAZOP). Finally, the target Safety Integrity Level (SIL) of each software module is determined based on its criticality level and the overall safety objective (i. e., SIL) of the system it resides in. As an example, this proposed method is applied to a nuclear power plant safety-critical system to demonstrate the detail application process and to verify the feasibility of the method. Compared with the existing software criticality analysis methods, this method has better operability and verifiability, and can be utilized as a technical guidance for the software criticality analysis of nuclear power plant digital control systems.

scholarly journals Probabilistic safety assessment (PSA) of a safety critical system: a case study of a nuclear power plant

2018 ◽  
Vol 7 (2.12) ◽  
pp. 210
Author(s):  
Vinay Kumar ◽  
Dewanshu Pratihar ◽  
Anil Kumar Tripathi
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Safety Assessment ◽  
Critical System ◽  
Probabilistic Safety Assessment ◽  
Safety Critical ◽  
Safety Critical System ◽  
Wide Range ◽  
Probabilistic Safety

Probabilistic Safety Assessment approach has been successfully applied in engineering, economics, computer science and statistics to re-solve a wide range of safety-related problems. However, using Probabilistic Safety Assessment for quantifying safety of a safety critical system is a challenging task in Safety Engineering community. This method plays an essential role in analyzing safety of safety critical sys-tems and its various components. Therefore, in this paper, we present Probabilistic Safety Assessment framework which can be used to quantify the critical failures of a systems. The approach is well demonstrated on a Digital Feed Water Control System uses in a Nuclear Power Plant as safety critical system. 

Design of the Communication Independence for ACPR1000 Nuclear Power Plant Digital Safety I&C System

2017 ◽  
Author(s):  
Sun Na ◽  
Shi Gui-lian ◽  
Xie Yi-qin ◽  
Li Gang ◽  
Jiang Guo-jin
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Communication Networks ◽  
Nuclear Power ◽  
Protection System ◽  
Plant Safety ◽  
Safety And Reliability ◽  
Digital Protection ◽  
Architecture And Design ◽  
Reactor Protection System

Communication independence is one of the key criteria of digital safety I&C system design. This paper mainly analyzes the requirements for communication independence in safety regulations and standards, and then introduces the architecture and design features, including communication failure processing measures, of communication networks of ACPR1000 nuclear power plant safety digital protection system based on FirmSys platform developed by CTEC. The communication design meets the regulations requirements and effectively improves the safety and reliability of the system, and it is successfully applied in reactor protection system (RPS) of Yang Jiang nuclear power plant unit 5&6. In addition this design can provide reference for communication designs of other NPPs and industries.

An Evaluation of Nuclear Power Plant Safety Parameter Display Systems

1981 ◽  
Vol 25 (1) ◽  
pp. 110-114 ◽  
Author(s):  
David D. Woods ◽  
John A. Wise ◽  
Lewis F. Hanes
Keyword(s):  
Power Plant ◽  
Nuclear Power Plant ◽  
Nuclear Power ◽  
Evaluation Tool ◽  
Training Simulator ◽  
Plant Safety ◽  
Power Plant Control ◽  
Safety Parameter ◽  
Nuclear Power Plant Safety ◽  
Display Systems

Two proposed safety parameter display systems, of the type to be required in nuclear power plant control rooms, were evaluated using a training simulator and experienced crews undergoing refresher training. A decision analysis approach was used. The discussion addresses the effectiveness of the training situation as an evaluation tool and methodological issues.

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